NUREG-1650, Rev. 2 

 

The United States of America 

Fourth National Report for the 

Convention on Nuclear Safety 

September 2007 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 
 NUREG-1650, Rev. 2 

 

The United States of America 

Fourth National Report for the 

Convention on Nuclear Safety 

Manuscript Completed: September 2007 

Date Published: September 2007 

 

September 2007 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 
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NUREG-1650, Rev. 2

The United States of America

Fourth National Report for the

Convention on Nuclear Safety

Manuscript Completed: September 2007

Date Published: September 2007

September 2007

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

ABSTRACT

The United States Nuclear Regulatory Commission has Updated NUREG-1650, Rev. 1, "The

United States Third National Report for the Convention on Nuclear Safety," issued September

2004, and will submit this report for peer review at the fourth review meeting of the Convention

on Nuclear Safety at the International Atomic Energy Agency in April 2008. This report

addresses the safety of land-based commercial nuclear power plants in the U.S. 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, priority 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. New to this update is

a contribution by the Institute of Nuclear Power Operations (INPO) describing work done by the

U.S. nuclear industry to ensure safety. Since the prime responsibility for the safety of a nuclear

installation rests with the license holder, it was decided to involve the nuclear industry

.collectively through INPO, in explaining how the nuclear industry maintains and improves

nuclear safety.

Paperwork Reduction Act Statement

This NUREG does not contain any information collection requirements and, therefore, is not

subject to the requirements of the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).

Public Protection Notification

The NRC may not conduct or sponsor, and a person is not required to respond to a request for

information or an information collection requirement unless the requesting document displays a

currently valid OMB control number.

iii

CONTENTS

A B S T R A C T .................................................... ; .......... iii

EXECUTIVE SUMMARY. ................................. xi

P A R T 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . 1

INTR O D U C T IO N ........................................................... 3

Purpose and Structure of this Report ...................................... 3

Summary of Changes to the Fourth U.S. National Report ....................... 5

The U.S. National Policy toward Nuclear Activities ........................... 12

National Nuclear Program s ............................................. 12

Power Uprate Program .......................................... 12

New Reactor Licensing ..................... ..................... 13

Reactor Oversight Process ....................................... 14

License Renew al ............................................... 14

Survey of Current Safety Issues ......................................... 15

Reactor Materials Degradation Issues ............................... 15

Unanticipated Equipment Problems from Power Uprates ................. 16

PWR Post-LOCA Chemical Formation ............................. 17

Status of Safety Issues Discussed in Third U.S. National Report ................ 19

Reactor Materials Degradation Issues ............................. 19

PWR Containment Sump Performance ............................ 19

Electric Grid Reliability ........................................ 20

Emergency Preparedness and Security ............................ 20

Other Maior Regulatory Accomplishments ............................... 22

Issuance of Early Site Permits for the Clinton and Grand Gulf Sites ........ 22

Design-Basis Threat ............................................ 22

AP1000 Design Certification ...................................... 22

Risk-Informed Categorization and Treatment of SSCs for Nuclear Power

R eactors .................................................. 22

CONCLUSIONS ON THE U.S. NATIONAL REPORT FROM THE

THIRD REVIEW MEETING ..... . ... ................................ 25

Items Resulting from Country Group Session ............................... 25

NRC Areas for Improvement and Maior Challenges for the Future ............... 26

NRC Maior Management Challenges for the Future .......................... 27

P A R T 2 .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 9

ARTICLE 6. EXISTING NUCLEAR INSTALLATIONS ............................. 31

6.1 Introduction .................................................. 31

6.2 Nuclear Installations in the United States ................................ 32

6.3 Regulatory Processes and Programs .................................. 32

6.3.1 Reactor Licensing .......................................... 32

6.3.2 Reactor Oversight Process .................................. 32

6.3.3 Industry Trends Program .................................... 34

V
6.3.4 Accident Sequence Precursor Program ....................... 34

6.3.5 Operating Experience Program ................................ 35

6.3.6 Program for Resolving Generic Issues .......................... 36

6.3.7 R ulem aking ........................ . ...................... 36

6.'3'8 Fire Regulation Program .................................... 37

6.3.9 Decom m issioning .......................................... 37

6.3.10 Reactor Safety Research Program ........................... 38

6.3.11 Special Programs for Public Participation ....................... 38

ARTICLE 7. LEGISLATIVE AND REGULATORY FRAMEWORK ..................... 41

7.1 Legislative and Regulatory Framework ................................. 41

7.2 Provisions of the Legislative and Regulatory Framework .................. 41

7.2.1 National Safety Requirements and Regulations ................... 41

7.2.2 Licensing of Nuclear Installations .............................. 42

7.2.3 Inspection and Assessment .................................. 43

7.2.4 Enforcem ent ............................................. 43

ARTICLE 8. REGULATORY BODY ............................................ 45

8.1 The Regulatory Body ........................................... 45

8.1.1 M andate ........... ...................................... 45

8.1.2 Authority and Responsibilities ................................ 45

8.1.3 Structure of the Regulatory Body .............................. 46

8.1.4 International Responsibilities and Activities ...................... 49

8.1.5 Financial and Human Resources ............................ 52

8.1.6 Position of the NRC in the Governmental Structure ............... 55

8.1.7 Report of the IRRS Self-Assessment Team ....................... 58

8.2 Separation of Functions of the Regulatory Body from Those of Bodies

Prom oting Nuclear Energ ......................................... 58

ARTICLE 9. RESPONSIBILITY OF THE LICENSE HOLDER ........................ 61

9.1 Introduction ................... ........ .......................... 6 1

9.2 The Licensee's Prime Responsibility for Safety .......................... 61

9.3 NRC Enforcement Program ......................................... 62

ARTICLE 10. PRIORITY TO SAFETY .......................................... 65

10.1 Background .................................................... 65

10.2 Probabilistic Risk Assessment Policy ................................. 65

10.3 Applications of Probabilistic Risk Assessment ......................... 66

10.3.1 Risk-Informed Inservice Testing .............................. 66

10.3.2 Risk-Informed Inservice Inspection ............................ 67

10.3.3 Risk-Informed Technical Specification Changes ................. 67

10.3.4 Development of Standards .................................. 68

10.4 Safety Culture ................ ........... .......... .... ..... 68

10.4.1 NRC Monitoring of Licensee Safety Culture ..................... 68

10.4.2 The NRC Safety Culture ................................. 71

ARTICLE 11. FINANCIAL AND HUMAN RESOURCES ............................. 73

11.1 Financial Resources .............................................. 73

vi
11.1.1 Financial Qualifications Program for Construction and Operations . .. 74

11.1.2 Financial Qualifications Program for Decommissioning ............ 75

11.1.3 Financial ProtectionProgram for Liability Claims Arising

from Accidents ........................................... 75

11.1.4 Insurance Program for Onsite Property Damages Arising

from Accidents ................. ....................... 76

11.2 Regulatory Requirements for Qualifying, Training,

and Retraining Personnel ...................................... 76

11.2.1 Governing Documents and Process ........................... 76

11.2.2 Experience ........................................... 78

ARTICLE 12. HUMAN FACTORS ............................................. 79

12.1 Goals and Mission of the Program ................................. 79

12.2 Program Elements ............................................ 79

12.3 Significant Regulatory Activities ..................................... 80

12.3.1 Human Factors Engineering Issues ........................... 80

12.3.2 Emergency Operating Procedures and Plant Procedures .......... 81

12.3.3 Shift Staffing ............................................ 81

12.3.4 Fitness-for-Duty ....................................... 82

12.3.5 Human Factors Information System ......................... 83

12.3.6 Support to Event Investigations and For-Cause

Inspections and Training ................................. 83

ARTICLE 13. QUALITY ASSURANCE ........................................... 85

13.1 Background ................................................. 85

13.2 Regulatory Policy and Requirements ............................... 85

13.2.1 Appendix A to 10 CFR Part 50 ............................. 86

13.2.2 Appendix B to 10 CFR Part 50 ............................. 86

13.2.3 Approaches for Adopting More Widely Accepted

International Quality Standards .............................. 86

13.3 QA Regulatory Guidance ........................................... 86

13.3.1 Guidance for Design and Construction Activities ................. 87

13.3.2 Guidance for Operational Activities ........................... 87

13.3.3 Guidance for Staff Reviews for Licensing ..................... 87

13.4 Q A Program s .................................................... 87

ARTICLE 14. ASSESSMENT AND VERIFICATION OF SAFETY ..................... 89

14.1 Ensuring Safety Assessments throughout Plant Life ..................... 89

14.1.1 Maintaining the Licensing Basis ............................ 89

14.1.2 License Renew al ......................................... 92

14.1.3 The United States and Periodic Safety Reviews ................. 94-

14.2 Verification by Analysis, Surveillance, Testing, and Inspection ............. 103

ARTICLE 15. RADIATION PROTECTION ........ .. ............................ 105

15.1 Authorities and Principles .......................................... 105

15.2 Regulatory Framework ........................................ 106

15.3 R egulations ................................................... 107

15.4 Radiation Protection Activities .................. .................... 108

vii
15.4.1 Control of Radiation Exposure of Occupational Workers .......... 108

15.4.2 Control of Radiation Exposure of Members of the Public .......... 109

ARTICLE 16. EMERGENCY PREPAREDNESS ............................... 111

16.1 Background 11.........................

16.2 Offsite Emergency Planning and Preparedness ........................ 112

16.3 Emergency Classification System and Emergency Action Levels ........... 113

16.4 Recommendations for Protective Action in Severe Accidents .............. 1,13.

16.5 Inspection Practices-Reactor Oversight Process for

Emergency Preparedness ......................................... 114

16.6 Responding to an Emergency ..................................... 116

16.6.1 Federal Response ..................................... 116

16.6.2 Licensee, State, and Local Response ........................ 117

16.6'.3 The NRC's Response ....................... .......... 117

16.6.4 Security Aspects Supporting Response ....................... 118

16.7 International Arrangements ..................................... 118

A RT IC LE 17. SITING ...................................................... 121

17.1 Background ................................................ 121

17.2 Safety Elem ents of Siting ......................................... 122

17.2.1 Background ............................. 122

17.2.2 Assessments of Seismic and Geological Aspects of Siting ........ 123

17.2.3 Assessments of Radiological Consequences ................... 123

17.3 Environmental Protection Elements of Siting ........................... 125

17.3.1 Governing Documents and Process .......................... 125

17.3.2 Other Considerations for Siting Reviews ...................... 126

ARTICLE 18. DESIGN AND CONSTRUCTION .................................. 127

18.1 Defense-in-Depth Philosophy ................................... 127

18.1.1 Governing Documents and Process........................... 127

18.1.2 Experience ............................................. 129

18.2 Technologies Proven by Experience or Qualified by Testing or Analysis ..... 130

18.3 Desigqn for Reliable, Stable, and Easily Manageable Operation ............ 130

18.3.1 Governing Documents and Process ........................ 130

18.3.2 Experience ............................................ 131

ARTICLE 19. OPERATION ........................ ..... ,................. 133

19.1 Initial Authorization to Operate ...................................... 133

19.2 Definition and Revision of Operational Limits and Conditions .............. 135

19.3 Approved Procedures ............ I. ........................ 136

19.4 Procedures for Responding to Anticipated Operational Occurrences

and Accidents ............................................. 136

19.5 Availability of Engineering and Technical Support ..................... 136

19.6 Incident Reporting ........................................... 137

19.7 Programs to Collect and Analyze Operating Experience ................. 137

19.8 Radioactive W aste .............................................. 139

APPENDIX A NRC STRATEGIC PLAN 2004-2009 ............................. 141

Viii
APPENDIX B NRC MAJOR MANAGEMENT CHALLENGES FOR THE FUTURE ..... ;.. 145

APPENDIX C REFERENCES ............................................... 151

APPENDIX D ABBREVIATIONS ............................................. 159

APPENDIX E ACKNOWLEDGMENTS ........................................ 161

ANNEX 1 U.S. COMMERCIAL NUCLEAR POWER REACTORS ..................... 163

PA R T 3 ................ ................................................. 175

CONVENTION ON NUCLEAR SAFETY REPORT: THE ROLE OF THE INSTITUTE OF

NUCLEAR POWER OPERATIONS IN SUPPORTING THE UNITED STATES COMMERCIAL

NUCLEAR ELECTRIC UTILITY INDUSTRY'S FOCUS ON NUCLEAR SAFETY

ix

EXECUTIVE SUMMARY

This is the "United States Fourth National Report for the Convention on Nuclear Safety"

(NUREG-1650, Rev. 2). The third report was issued by the United States (U.S.) Nuclear

Regulatory Commission (NRC) in September 2004. This report will be submitted for peer

review at the fourth review meeting of the Convention on Nuclear Safety, to be convened at the

International Atomic Energy Agency in April 2008. This report addresses the safety of landbased commercial nuclear power plants in the U.S. 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, priority 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 in the peer review of the third review meeting in April

2005 and discusses challenges and issues that have arisen since that time. The NRC's

challenges identified in the third review meeting were:

(1) ensuring that the post-September 11, 2001, emphasis on security does not have an

adverse effect on operational safety,

(2) continuing the focus on safe long-term operation and emerging technical issues,

(3) preparing for the potential licensing of new reactors,

(4) appropriately emphasizing safety culture, safety management, and organizational

aspects of nuclear safety,

(5) continued monitoring and analysis of operating experience and timely implementation of

lessons learned, and

(6) expanding risk-informed, performance-based concepts while ensuring a consistent

regulatory approach.

At the third review meeting, highlighted planned initiatives included:

(1) enhancing the Reactor Oversight Process to increase NRC inspector review of licensee

safety-conscious work environment,

_(2) working with the international community to define the ways and means to detect

deteriorating performance, including safety culture and safety management, and

(3) planning for the performance of an International Regulatory Review Team (renamed the

Integrated Regulatory Review Service) self-assessment.

This report updates the status of these initiatives.

The safety issues discussed in the third U.S. National Report were reactor materials problems,

pressurized-water reactor containment sump performance, electric grid reliability, emergency

preparedness, and security. The status of these safety issues and those that have arisen since

2004 is discussed in this report, namely, reactor materials degradation problems, and

pressurized-water reactor containment sump blockage due to post-loss-of-coolant accident

chemical formation, and issues regarding post-power-uprate equipment.

xi
Since new reactors are of increasing international interest, this report also discusses the NRC's

preparations for licensing new reactors in the United States, an activity the NRC has not

undertaken in the past 30 years.

The NRC considers managing human capital a major focus area, and this report discusses how

the agency recruits and trains new staff, captures and passes on lessons learned from past

events, and attempts to capture and pass on the knowledge of long-term staff as they prepare

to retire.

For the first time, the Institute of Nuclear Power Operations (INPO) is providing input to the U.S.

National Report. Since the prime responsibility for the safety of a nuclear installation rests with

the license holder, it was decided to involve the nuclear industry collectively through INPO, in

explaining how the nuclear industry maintains and improves nuclear safety.

xii
PART 1

1

INTRODUCTION

This section describes the purpose and structure of the "United States Fourth National Report

for the Convention on Nuclear Safety," the United States (U.S.) national policy toward nuclear

activities, the main national nuclear programs, and the current nuclear safety issues. It also

highlights major regulatory accomplishments since the previous (third) U.S. National Report

was submitted in 2004 (see NUREG-1650, Rev. 1, "The United States Third National Report for

the Convention on Nuclear Safety," September 2004).

Purpose and Structure of this Report

The United States of America is submitting this updated report for peer review to the Fourth

Review Meeting of the Contracting Parties to the Convention on Nuclear Safety (the

Convention). 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 in order 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 U.S. Fourth National

Report, principally using agency information that is publicly available. This updated report

follows the format of the "U.S. Third National Report for the Convention on Nuclear Safety," and

is designed-to be a.stand-a0lne document. Hence, this report duplicates some of the

information presented in the 2004 (third) report. To facilitate peer review; this report includes a

summary table of the main changes to the report. Also new is a contribution by the Institute of

Nuclear Power Operations (INPO) describing work done by the U.S. nuclear industry to ensure

safety (see Part 3). INPO is a non-governmental corporation founded in 1979 by the U.S.

nuclear industry to collectively promote the highest levels of safety and reliability of U. S.

nuclear plants. Since the prime responsibility for the safety of a nuclear installation rests with

the license holder, it was decided to involve the nuclear industry collectively through INPO, in

explaining how the nuclear industry maintains and improves nuclear safety.

"Commission" may also refer to the Chairman and Commissioners who head the NRC.

3
Following the introduction, Part I of the report continues with a summary of changes to the

report (Table 1), to make it easier for contracting parties to review sections of interest to them.

This table is followed by a section on the conclusions of the previous review meeting. Part 2

discusses Convention articles 6-19 and includes an annex and references to provide more

detailed information as appropriate. 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 by the chapter, and a discussion of how the United States

meets the obligations of the article. Articles 6-9 summarize the legislative and regulatory

system governing the safety of nuclear installations and discuss the adequacy and

effectiveness of that system. Articles 10-16 address general safety considerations and

summarize major safety-related features. Articles 17-19 address the safety of installations.

Part 2 concludes with a series of appendices that discuss the NRC's main challenges as

described in the NRC Strategic Plan and the 2006 Inspector, General's report, followed by

appendices of references, abbreviations, and acknowledgments. Annex 1 of the report is a list

of U. S. nuclear plants. Part 3 of the report is a new section that contains INPO's contribution.

This report does not explicitly discuss Articles 1-5. These articles are fulfilled throughout the

report, and by the existence of the report, itself. In accordance with Article 1, the report ..

illustrates how the U.S. Government meets the objectives of the Convention. The report also

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 of Article 5 on reporting. In addition, the information in this report is available in

more detail at the NRC's public Web site.

4
Summary of Changes to the Fourth U.S. National Report

To facilitate peer review of this report, the following table summarizes the changes from the

third U.S. National Report. Changes from the third report are identified by a revision bar along

the left margin of the page.

Table 1 Summary of Changes to the Fourth U.S. National Report

Section of Report Change

PART 1

INTRODUCTION Updated.

Purpose and Structure of this Report Added table of main changes and discussed modified

structure, including the contribution by INPO.

The U.S. National Policy toward Nuclear Activities No change.

National Nuclear Programs

Power Uprate.Program Shortened description.

New Reactor Licensing Updated number of design certifications and designs

under review; added discussion of early site permits and

of a new construction inspection organization.

Reactor Oversight Process Updated to incorporate 2006 assessment results of the

*now 7-year old process.

License Renewal Updated number of renewals. Shortened discussion.

Survey of Main Current Safety Issues Updated status/resolution of 2004 issues and added new

issues.

Reactor Materials Degradation Issues Added new safety issues.

Unanticipated Equipment Problems from Power Updated as safety issue.

Uprates

PWR Post-LOCA Chemical Formation Added as new safety issue.

Status of Safety Issues Discussed in Third U.S. Added.

National Report

Reactor Materials Degradation Issues Updated status.

PWR Containment Sump Performance Discussed above under PWR Post-LOCA Chemical

Formation.

Electric Grid Reliability Updated status.'

Emergency Preparedness and Security Updated status.

Other Major Regulatory Accomplishments Updated to discuss early site permits, regulation involving

design-basis threats, the AP1000 design certification, and

risk-informed categorization and treatment of structures,

systems, and components for nuclear power reactors.

Nuclear Installations in the United States (Table) Removed because it duplicates Section 6.1.

5
Section of Report Change

CONCLUSIONS ON THE U.S. NATIONAL

REPORT FROM THIRD REVIEW MEETING New.

Items Resulting from Country Group Session New.

NRC Areas for Improvement and Major No change in 2004-2009 NRC Strategic Plan.

Challenges for the Future

NRC Major Management Challenges for the Updated to incorporate 2006 Inspector General report.

Future Eight of the nine challenges are essentially the same as in

the previous U.S. National Report. In 2006, the Inspector

General identified a new challenge titled, Ability to Meet

the Demand for Licensing New Reactors and removed a

challenge titled, Intraagency Communication.

PART 2: RESPONSE TO ARTICLES OF THE

CONVENTION

ARTICLE 6. EXISTING NUCLEAR

INSTALLATIONS

6.1 Introduction New. Renumbered subsequent subsections.

6.2 Nuclear Installations in the United States Updated to current status.

6.3 Regulatory Processes and Programs

6.3.1 Reactor Licensing Updated expectations about early site permits and design

certification applications.

Current power uprate issues are discussed under Survey

of Main Current Safety Issues.

6.3.2 Reactor Oversight Process Condensed section and substituted new material.

6.3.3 Industry Trends Program Updated current experience.

6.3.4 Accident Sequence Precursor Program Updated current experience and added some details on

changes in the program.

6.3.5 Operating Experience Program Added.

6.3.6 Program for Resolving Generic Issues Updated NUREG-0933 to reference current revision

.. _ _ ._ __ ..... _ _ (October 2006).

6.3.7 Rulemaking No change.

6.3.8 Fire Regulation Program Added.

6.3.9 Decommissioning Shortened description and added early consideration of

decommissioning.

6.3.10 Research Program Added work regarding new reactors.

6.3.11 Programs for Public Participation, Handling Moved discussion of Differing Professional Opinions;

Petitions, and Allegations made minor changes to allegation discussion. Added

description of petition process for rulemaking.

ARTICLE 7. LEGISLATIVE AND REGULATORY

FRAMEWORK

6
Section of Report Change

7.1 Legislative and Regulatory Framework No change.

7.2 Provisions of the Legislative and Regulatory

Framework

'7.2.1 National Safety Requirements and Shortened discussion.

Regulations

7.2.2 Licensing of Nuclear Installations Revised discussion of Part 52.

7.2.3 Inspection and Assessment No change.

7.2.4 Enforcement Updated the amount adjusted by the Federal Civil

Penalties Inflation Adjustment Act (currently $130,000).

ARTICLE 8. REGULATORY BODY Shortened descriptions of offices.

8.1 The Regulatory Body Reorganized and renumbered subsequent subsections.

8.1.1 Mandate No change.

8.1.2 Authority and Responsibilities Reorganized to include discussion of NRC as an

independent agency.

8.1.3 Structure of the Regulatory Body Noted organizational changes.

8.1.4 International Responsibilities and Activities Updated international activities.

8.1.5 Financial and Human Resources Updated budget information. Added discussion of human

resources regarding workforce planning and knowledge

management.

8.1.6 Position of the NRC in the Governmental Added Department of Transportation.

Structure Updated.

8.1.7 Report of the Integrated Regulatory Review New.

Service Self-Assessment Team

8.2 Separation of Functions of the Regulatory Shortened.

Body from Those of Bodies Promoting Nuclear

Energy

ARTICLE 9. RESPONSIBILITY OF THE LICENSE

,HOLDER

9.1 Introduction No change.

9.2 The Licensee's Prime Responsibility for No change.

Safety

9.3 The NRC Enforcement Program Updated debt collection dollar amount.

Added Alternative Dispute Resolution Program.

Updated experience.

ARTICLE 10. PRIORITY TO SAFETY

10.1 Background Replaced previous examples of risk-informed rules with

current ones. Added discussion of proposed set of riskinformed, performance-based requirements (10 CFR Part

53).

7
Section of Report Change

10.2 Probabilistic Risk Assessment Policy Shortened.

10.3 Applications of Probabilistic Risk Added discussion of a risk-informed, performance-based

Assessment plan and deleted discussion of the Risk-Informed

Regulation Implementation Plan. Deleted previous

discussions and substituted with Recent Developments.

10.4 Safety Culture

10.4.1 NRC Monitoring of Licensee Safety Culture Moved and thoroughly revised.

10.4.2 NRC Safety Culture Thoroughly revised.

ARTICLE 11. FINANCIAL AND HUMAN

RESOURCES

11.1 Financial Resources Updated to incorporate changes in the dollar amounts for

liability under the Price-Anderson Act of 1957. Condensed

section.

11.2 Regulatory Requirements for Qualifying, Thoroughly revised. Shortened discussions.

Training, and Retraining Personnel

ARTICLE 12. HUMAN FACTORS Updated the program on human performance,

documents, issues, and experience.

12.1 Goals and Mission of the Program Portions moved to new 12.2 and 12.3.

12.2. The NRC Program on Human Performance Updated the program elements.

12.3 Significant Regulatory Activities Expanded discussion of fitness-for-duty.

Thoroughly revised discussion of Human Factors

Information System.

Updated experience.

12.2 Significant Research Activities Deleted.

ARTICLE 13. QUALITY ASSURANCE Condensed section. Added a discussion of the

construction inspection program for expected new

reactors.

ARTICLE 14. ASSESSMENT AND

VERIFICATION OF SAFETY

14.1 Ensuring Safety Assessments throughout Updated the Brown's Ferry restart.

Plant Life

14.1.1 Maintaining the Licensing Basis Added discussion of amendment process.

14.1.2 License Renewal Minor editorial change.

14.1.3 The United States and Periodic Safety No change.

Reviews

14.2 Verification by Analysis, Surveillance, Added paragraph on license conditions (10 CFR 50.54f).

Testing and Inspection

ARTICLE 15. RADIATION PROTECTION

15.1 Authorities and Principles Updated work on 2007 ICRP recommendations.

8
Section of Report Change

15.2 Regulatory Framework Minor editorial changes.

15.3 Regulations Minor edits and clarifications.

15.4ý Radiation Protection Activities. Updated dose rates; expanded the discussion of Appendix

I to 10 CFR Part 20. Added clarifications.

15.4.1 Control of Radiation Exposure of Updated dose information.

Occupational Workers

15.4.2 Control of Radiation Exposure of Members Added ground water contamination issue.

of the Public

ARTICLE 16. EMERGENCY PREPAREDNESS

16.1 Background No change.

16.2 Offsite Emergency Planning and Updated section to change references to FEMA to

Preparedness DHS/FEMA.

16.3 Emergency Classification System and No change.

Emergency Action Levels

16.4 Recommendations for Protective Action in Added sentence on generic communications. States

Severe Accidents changes in potassium iodide considerations.

16.5 Inspection Practices-Reactor Oversight Changed the names of areas to be inspected.

Process for Emergency Preparedness

16.6 Responding to an Emergency Discussed changes resulting from the response to the

terrorist events of September 11, 2001, and Hurricane

Katrina. Updated security requirements.

16.7 International Arrangements No change.

ARTICLE 17. SITING

17.1 Background Updated status of early site permits and siting

applications.

17,2 Safety Elements of Siting

17.2.1 Background Added new developments in guidance.

17.2.2 Assessments of Seismic and Geological Added new developments in seismic hazard assessments.

Aspects of Siting

17.2.3 Assessments of Radiological Updated experience in the use of an alternative source

Consequences term.

17.3 Environmental Protection Elements of Siting

17.3.1 Governing Documents and Process Added some details on site approval process.

17.3.2 Other Considerations for Siting Reviews Cited updated guidance.

ARTICLE 18. DESIGN AND CONSTRUCTION

18.1 Defense-in-Depth Philosophy

9
Section of Report Change

18.1.1 Governing Documents and Process No change.

18.1.2 Experience Updated the reactivation of Watts Bar and the designs

certified and under review.

18.2 Technologies Proven by Experience or No change.

Qualified by Testing or Analysis

18.3 Design for Reliable, Stable, and Easily

Manageable Operation

18.3.1 Governing Documents and Process Added governing documents.

18.3.2 Experience Added experience, digital instrumentation and control,,

and cyber security.

ARTICLE 19. OPERATION

19.1 Initial Authorization to Operate No change.

19.1.1 Governing Documents and Process No change. Now 19.1.

19.1.2 Experience Deleted.

19.2 Operational Limits and Conditions Are Shortened.

Defined and Revised

19.3 Approved Procedures Shortened.

19.4 Procedures for Responding to Anticipated Shortened.

Operational Occurrences and Accidents

19.5 Availability of Engineering and Technical No change.

Support

19.6 Incident Reporting Expanded.

19.7 Programs to Collect and Analyze Operating Expanded and revised to discuss revised Operating

Experience Experience Program.

19.8 Radioactive Waste - Updated to cite most recent report for the Joint Convention

on the Safety of Spent Fuel Management and on the

Safety of Radioactive Waste Management.

Appendix A: NRC Strategic Plan 2004-2009 No change. The NRC is currently developing a new strategic plan.

Appendix B: NRC Major Management Challenges Updated.

For The Future

Appendix C: References Updated.

Appendix D: Abbreviations Updated.

Appendix E: Acknowledgments Updated.

ANNEX 1: U.S. Commercial Nuclear Power Updated.

Reactors

10
Section of Report Change

PART 3: ROLE OF THE INSTITUTE OF New.

NUCLEAR POWER OPERATIONS IN

SUPPORTING THE U.S. COMMERCIAL

NUCLEAR ELECTRIC UTILITY INDUSTRY'S

FOCUS ON NUCLEAR SAFETY

11
The U.S. National Policy toward Nuclear Activities

The Energy Reorganization Act of 1974 created the 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. The

agency also has a role in combating the proliferation, of nuclear materials worldwide. 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.

The NRC views nuclear regulation as the public's business and, as such, it must be transacted

openly and candidly to maintain the public's confidence. 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 certain proprietary business

material, facility safeguards information, sensitive pre-decisional information, and information

supplied by foreign countries that is deemed to be sensitive, the NRC makes the documentation

that it uses in its decision-making available in the agency's Public Document Room in Rockville,

Maryland, and on the agency's public Web site at http://www.nrc.cov. As a result, a significant

amount of information about nuclear activities and the national policy regarding them is available

to everyone.

The NRC's interpretation of regulations continues to evolve from a prescriptive, deterministic

approach toward a more risk-informed and performance-based regulatory approach. Improved

probabilistic risk assessment (PRA) techniques, combined with more than four decades of

accumulated experience with operating nuclear power reactors, led the Commission to revise or

eliminate certain requirements. The Commission is also prepared to strengthen the regulatory

system when risk considerations reveal the need.

National Nuclear Programs

The NRC has a number of programs and processes to protect the health and safety of the public

and the environment and to meet the obligations of the Convention. Key programs and

processes in the reactor arena comprise a well-established licensing process, which includes

power uprates, new reactor licensing and early site permits, reactor oversight, and license

renewal. For more details on these programs, see the section of this report that pertains-to

Article 6.

Power Uprate Program

Under its licensing program, the NRC carefully reviews requests to raise the maximum thermal

power level at which a plant may be operated, known as power uprate requests. The focus of

the NRC review of these requests has been, and will continue to be, on safety. The agency

closely monitors operating experience to identify safety issues that may affect the

implementation of power uprates.

12
Power uprates can be classified as (1) measurement uncertainty recapture power uprates, (2)

stretch power uprates, and (3) extended power uprates (EPUs). Measurement uncertainty

recapture power uprates are less than a two percent increase and are achieved by implementing

enhanced techniques for calculating reactor power. Stretch power uprates are typically up to an

eight percent increase and are generally within the design capacity of the plant. Stretch power

uprates usually involve changes to instrumentation setpoints and do not generally involve major

plant modifications. EPUs are usually greater than stretch power uprates and require significant

modifications to' major balance-of-plant equipment. The NRC has approved EPUs of Up to

20 percent.

Power uprate issues are discussed under the heading, Survey of Main Current Safety Issues.

New Reactor Licensing

The industry has expressed interest in constructing new nuclear power plants in the United,

States and indicated as of June 2007, that it may submit applications for up to 28 new reactor

licenses over the next few years. The NRC is ready to accept these applications. New nuclear

power plants will likely use the licensing process specified in the recently revised Title 10,

Part 52, "Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear

Power Plants," of the Code of Federal Regulations (10 CFR Part 52), which is designed to be

more stable and predictable than the process specified in 10 CFR Part 50, "Domestic Licensing

of Production and Utilization Facilities." This licensing process resolves all safety and

environmental issues, as well as emergency preparedness and security, before a new nuclear

power plant is constructed.

The Commission has certified four new reactor designs under 10 CFR Part 52, making them

readily available for referencing in new plant licensing submittals. These designs are General

Electric's Advanced Boiling-Water Reactor, and Westinghouse's AP600, AP1 000, and

System 80+ (designed and licensed by Combustion Engineering).

In addition to the four advanced reactor designs that have been certified, staff is currently

reviewing General Electric's economic simplified boiling-water reactor (ESBWR) design

certification application. Furthermore, the NRC is preparing to receive and review two additional

design certification applications in the coming year.

By certifying nuclear reactor designs, the NRC resolves safety issues in a design certification

rulemaking. When an applicant submits an application for construction of a new nuclear power

plant using one of the certified designs, the license application review can proceed more

efficiently in a manner that ensures safety while minimizing unnecessary regulatory burden and

delays.

The NRC has received four early site permit applications for sites in Virginia, Illinois, Mississippi,

and Georgia. These sites currently host operating reactors. In March 2007, the Commission

approved the issuance of early site permits for the Illinois and Mississippi sites. These are the

first early site permits issued by the NRC and the first time this part of the 10 CFR Part 52

licensing process has been implemented. According to this process, environmental issues that

have been resolved in the early site permit proceedings cannot be re-opened during a combined

license proceeding.

13
In 2006, to better prepare the agency for the anticipated new reactor licensing and construction

inspection work, while ensuring that the agency maintains its focus on the safety and security of

currently operating reactors, the NRC established the Office of New Reactors. The agency also

established a dedicated Construction Inspection Organization in its Region II office in

Atlanta, Georgia that will carry out all construction inspection activities across the U.S., including

both the day-to-day onsite inspections and the specialized inspections needed to support NRC

oversight of the construction of new nuclear power plants.

There is one partially built plant, Watts Bar Nuclear Plant Unit 2, that had stopped construction

activities in the mid-1980s, and is planning to resume construction and pursue operating license

approval under 10 CFR Part 50. Watts Bar Unit 2 is a Westinghouse designed PWR located in

southeastern Tennessee and owned by the Tennessee Valley Authority.

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 advanced

reactor oversight. The NRC is participating in a multinational effort, the Multinational Design

Evaluation Program, to more efficiently review new reactor designs. The goal of this effort is to

make all new reactor reviews more safety-focused. NRC representatives are communicating

closely with representatives from the Finnish and French regulatory authorities concerning the

European Power Reactor designs that are under construction in Finland and slated to be

licensed in France and the-United States. A longer-term multinational effort is being undertaken

to establish reference regulatory practices and regulations for the review of current and future

reactor designs.

Reactor Oversight Process

The NRC's Reactor Oversight Process is now nearly 7 years old. In its annual self-assessment

for calendar year (CY) 2006, the NRC staff concluded that the Reactor Oversight Process

provided effective safety oversight. The Reactor Oversight Process is objective, risk informed,

understandable, and predictable and ensures safety, openness, and effectiveness. The NRC

has appropriately monitored operating nuclear power plant activities and focused agency

resources on performance issues in 2006, and plants Continue to receive a level of regulatory

oversight commensurate with their performance. The NRC staff expects to continually improve

the Reactor Oversight Process using lessons learned and recommendations from independent

evaluations. The NRC will also continue toactively solicit input from its internal'and external

stakeholders to further improve the Reactor Oversight Process.

License Renewal

The focus of the Commission's review of license renewal applications -is on maintaining plant

safety, with the primary emphasis on the effects of aging on important structures, systems,

and components (SSCs). 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. With the

improved 'economic conditions for operating nuclear power plants, the Commission has seen

sustained, strong interest in license renewal, which allows plants to operate up to 20 years

14
beyond their current operating licenses. TheAtomic Energy Act established the original 40-year

term, which was not based on technical limitations.

The decision to seek license renewal is.voluntary and rests entirely with nuclear power plant

owners. The decision is typically based on the plant's economic viability and whether it can

continue to meet the Commission's requirements. If the Commission approves all of the

applications that are currently under review, approximately one-half of the plants in the United

States will have had their operating licenses renewed. Based on statements from industry

representatives, the Commission expects nearly all sites to apply for license renewal.

Survey of Current Safety Issues

The NRC and its licensees currently face the following regulatory and safety issues:

reactor materials degradation issues

unanticipated equipment problems from power uprates

pressurized-water reactor (PWR) Emergency Core Cooling System (ECCS) sump

blockage due to post-loss-of-coolant accident (LOCA) chemical formation

Reactor Materials Degradation Issues

I Cases involving materials degradation include the Wolf Creek pressurizer dissimilar metal butt

welds, and the Duane Arnold jet pump riser safe end.

Wolf Creek Pressurizer Dissimilar Metal Butt Weld Cracking

In October 2006, Wolf Creek reported the presence of five circumferential indications in

dissimilar metal butt welds for pressurizer surge, safety, and relief nozzles, which was not

acceptable under Section XI of the American Society of Mechanical Engineers (ASME) Code.

The indications were determined to be flaws, most likely caused by PWSCC. The weld material

is Alloy 82/182, which is susceptible to PWSCC.

These flaws were of concern because they were circumferential and much larger than those that

had been previously discovered. Past indications have been axial or short circumferential flaws:

Also, this is the first time that multiple circumferential indications have been identified in a single

weld. This calls into question the assumptions and evaluations that the industry has made in

developing the timeliness guidelines for inspection of these welds.

The NRC staff analyzed the flaws in each of the three nozzle welds found at Wolf Creek to gain

insights as to when the flaws may have begun and to estimate how long it would have taken for

leakage or rupture to have occurred if the flaws had been left in service. All scenarios analyzed

for the surge nozzle flaw indicated that there was ample time, from the onset of leakage, for the

licensee to identify the leak and take appropriate action before failure of the weld. However, for

the flaws identified in the relief and safety nozzles, the analyses indicated that the flaws could

have caused the welds to fail in less than three years. A number of those analyses indicated,

that the failure could have occurred without any prior leakage, which would serve as a warning of

impending failure.

15
On the basis of the above information, the NRC issued confirmatory action letters to affected

licensees, asking them to confirm their commitment to inspect the pressurizer surge, spray,

safety, and relief nozzle welds by December 31, 2007. The letters asked to (1) implement

enhanced reactor coolant system leakage monitoring until the inspections are complete, (2)

repeat butt weld examinations every four years until the welds are either removed from service

or mitigated, and (3) report inspection results to the NRC.

Duane Arnold Jet Pump Riser Safe End Cracking Event

In February 2007, the licensee (Florida Power and Light) for Duane Arnold informed the NRC

that ultrasonic testing of an Inconel 82/182 weld between a low-alloy steel reactor vessel nozzle

and a stainless steel jet pump riser pipe safe end revealed evidence of a likely stress-corrosion

cracking flaw. Ultrasonic testing of eight Inconel 82/182 welds revealed evidence of two flaws in

welds between low-alloy steel reactor vessel nozzles and stainless steel jet pump riser pipe safe

ends. The piping at these locations is approximately one inch in wall thickness. Both

indications are located within the weld material, on the surface connected to the inside diameter

of the piping, and are approximately 55-75 percent through wall. Upon reevaluating data from

1999 and 2005 ultrasonic testing examinations, the licensee determined that these flaws had

been evident at the time of those examinations, but had -not been not identified. The licensee

repaired the welds with overlays. The plant-specific and generic implications of the Duane

Arnold information are now being evaluated.

Unanticipated Equipment Problems from Power Uprates

The NRC monitors operating experience at plants that have implemented power uprates. In

2002 and 2003, steam dryer cracking and flow-induced vibration damage on components and

supports for the main steam lines and feedwater lines occurred at some boiling water reactor

(BWR) plants. For example, the safety-related electromatic relief valves in the main steamlines

at Quad Cities Units 1 and 2 were damaged by excessive vibration during uprated power

operation (plants were uprated by up to 17.8 percent). Power upratings of more than eight

percent are defined by the NRC as EPUs. Also, the steam dryers at Quad Cities Units 1 and 2

developed cracks and, in some cases, fractured metal parts from the steam dryer fell into the

reactor pressure vessel and entered the steamlines leading to the turbine generator during EPU

operation. In addition, feedwater sampling probes at Dresden Units 2 and 3 broke loose within a

relatively short period of time under the higher feedwater flow conditions (authorized to 17

percent).

At EPU conditions, steamflow velocities can increase significantly. Plant experience has shown

that the higher main steamline flow can create an acoustic resonance in the steamlines as the

flow passes over branch lines. The acoustic resonance can create pressure waves that strike

the steam dryer in BWRs with sufficient force to cause the stressin the steam dryer to exceed

the, material fatigue limits. The acoustic resonance can also cause excessive vibration that

damages steamline components, such as relief valves and piping.

Therefore, the NRC has been performing more detailed reviews and inspections of plant

performance and power uprate license amendment requests with respect to adverse flow effects

on plant SSCs. However, the recognition that acoustic resonance can cause adverse flow

effects is relatively new to the nuclear power industry. The following discussion summarizes the

more significant plant operating experience related to adverse flow effects.

16
In response to the failure of the original steam dryers at Quad Cities Units 1 and 2 during EPU

operation, the licensee installed new, improved steam dryers, in May 2005. The Quad Cities

Unit 2 steam dryer design included pressure sensors, strain gages, and accelerometers to

monitor the loads on the steam dryer during restart to EPU conditions. The main steamlines on

both units were also instrumented to monitor loads during power ascension to EPU conditions.

Following the return to EPU operation in mid-2005, the licensee discovered significant

unexpected degradation of the actuators for several electromatic relief valves in the main

steamlines at Quad Cities Units 1 and 2. To reduce the acoustically-generated pressure

fluctuations and vibrations in the main steamlines, the licensee installed acoustic side branches

in the inlet lines of the electromatic relief valves and the main steam safety valves.

The original steam dryers in Dresden Units 2 and 3 were similar to the original Quad Cities

steam dryers, so they were subsequently modified to increase their structural capability. The

licensee initially operated the Dresden units at EPU conditions for several years without

significant damage. However, after discovering steam dryer damage at the Dresden units in

2005 and 2006, the licensee replaced the steam dryer in Dresden Unit 3-in November 2006.

The steam dryer in Dresden Unit 2 will be replaced in October 2007.

The NRC is applying lessons learned from operating experience and analysis of potential

adverse flow effects in reviewing power uprate requests for operating nuclear power plants and

design certification requests for new nuclear power plants. As part of this effort, the NRC has

updated relevant sections of NUREG-0800, "Standard Review Plan for the Review of Safety

Analysis Reports for Nuclear Power Plants", and Regulatory Guide 1.20, "Comprehensive

Vibration Assessment Program for Reactor Internals during Preoperational and Initial Startup

Testing," issued in March 2007, to further guide NRC reviewers and the nuclear industry to

evaluate potential adverse flow effects when considering power uprates in both operating and

new nuclear power plants.

PWR Post-LOCA Chemical Formation

The third U.S. National Report identified PWR containment sump performance as an issue.

That issue remains of concern, though substantial progress has been made in resolving it (e.g.,

by greatly enhancing strainer size and removing problematic materials). Consistent with NRC

Generic Letter (GL) 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation

during Design Basis Accidents at Pressurized-Water Reactors," dated September 13, 2004, the

NRC continues to target December 31, 2007, as the date for licensees to demonstrate that their

strainers will perform acceptably in the presence of expected plant-specific debris generation

and transport. However, one aspect of the issue, the potential for chemical effects on strainers

and downstream components, has turned out to be particularly challenging.

To address concerns about the potential for chemical precipitates and corrosion products to

significantly block a fiber bed and increase the head loss across an emergency core cooling

system sump screen, a joint NRC/industry Integrated Chemical Effects Testing Program was

started in 2004 and concluded in August 2005. The test program identified chemical

precipitation products, and followup testing and analyses were performed to address the effect

on head loss. On the basis of this effort, the NRC issued Information Notice (IN) 2005-26,

"Results of Chemical Effects Head Loss Tests in a Simulated PWR Sump Pool Environment," on

September 16, 2005.

17
The NRC conducted additional research to support evaluation efforts and provide-confirmatory

information. This work includes research on (1) chemical effects to determine whether the PWR

sump pool environment generates byproducts that contribute to sump clogging, (2) pump head

losses caused by the accumulation of containment materials and chemical byproducts, and (3)

the prediction of the chemical species that may form in these environments. Supplement 1 to IN

2005-26, "Additional Results of Chemical Effects Tests in a-Simulated PWR Sump Pool

Environment," dated January 20, 2006, provides additional information on test results related to

chemical effects in environments containing dissolved phosphate (e.g., from trisodium

phosphate) and dissolved calcium. The results discussed in the information notices clearly,

demonstrate that chemical effects can be significant. Follow-on testing sponsored by the

industry has also shown that these chemical effects tend to increase the potential for substantial

head loss.

The agency also conducted research on the transportability of coating chips in containment pool

environments and on the effect of ingested debris on downstream valve performance. In 2006,

the staff completed additional research on various aspects of sump clogging. All planned NRCsponsored research activities related to PWR sump clogging are now complete and

documented, though information that is obtained as the staff continues its reviews of industry

activities to close the issue may suggest the need for more NRC-sponsored research.

The NRC is currently reviewing an industry topical report that supports evaluation and testing of

chemical effects, and expects to receive another topical report addressing chemical effects

inside the reactor vessel. Licensees are developing protocols for integrated head-loss testing

that will include chemical effects. The NRC is monitoring this work, reviewing the protocols, and

observing and commenting on testing. Most of the testing is expected to occur in summer and

fall 2007. The staff will consider the test results in the licensees' final strainer designs, and in

demonstrations that their strainer designs are adequate. The NRC will use inputs from its review

of licensee responses to GL 2004-02, inspections, and audits of corrective actions by selected

licensees to support closing the PWR sump clogging issue, including the chemical effects.

18
Status of Safety Issues Discussed in Third U.S. National Report

Reactor Materials Degradation Issues

The reactor materials degradation issues in 2004 focused on PWSCC in PWR vessel upper and

lower head penetrations and other locations in the reactor coolant system and boric acid

corrosion. Cases discussed were a significant cavity in the reactor vessel head at the DavisBesse Nuclear Power Station and boron deposits around lower reactor vessel head penetration

nozzles at the South Texas Project Unit 1.

Davis-Besse Nuclear Power Station Reactor Vessel Head

In March 2002, the licensee for Davis-Besse Nuclear Power Station discovered a significant

cavity in the reactor vessel head. The cavity was next to a leaking nozzle with a through-wall

crack and in an area of the vessel head that had been covered with boric acid deposits

for several years. The NRC considers the Davis-Besse case as one of the most-significant in

the agency's recent history. Davis-Besse continues to draw much interest and comment from

the agency, the industry, all levels of government, and the public. Consequently, the NRC

continues to dedicate substantial effort to resolving both the technical and programmatic issues

that contributed to the degradation at this site. (For history and details, see the extensive

documentation regarding this case on the NRC public Web site at www.nrc.aov.)

South Texas Instrumentation Penetrations

Another significant case concerns lower reactor vessel head penetration nozzles. The lower.

head and bottom-mounted instrumentation penetrations of the South Texas Project,;Unit 1.,

reactor vessel were visually inspected on April 12, 2003, as a routine part of the unit's refueling

outage. The inspection found small amounts of white residue around 2 of the 58 penetrations at

the junction where the penetrations meet the lower reactor vessel head. The licensee performed

destructive examination of one of the penetrations and found a through-wall flaw resulting from

primary water stress-corrosion cracking (PWSCC). As a result-of these events, the NRC staff

issued Bulletin 2003-02, "Leakage from Reactor Pressure Vessel Lower -Head Penetrations and

Reactor Coolant Pressure Boundary Integrity," on August 21, 2003.

In September 2006, the NRC staff issued NUREG-1 863, "Review of Responses to NRC Bulletin

2003-02-Leakage from Reactor Pressure Vessel Lower Head Penetrations and Reactor

Coolant Pressure Boundary Integrity." The NUREG summarizes the available information on the

current and future inspection programs and the industry and agency's ongoing plans to ensure

the integrity of the penetrations. Even though there was no evidence of bottom-mounted

instrumentation penetration leakage at other plants, the NRC staff believes that continuous

monitoring of the penetrations is necessary to- ensure the integrity of the lower reactor vessel

head.

PWR Containment Sump Performance

For the status of this issue, see the discussion above under the heading, PWR Post-LOCA

Chemical Formation.

19
I Electric Grid Reliability

The long duration of the blackout in the eastern United States and Canada on August 14, 2003,

highlighted the need to further consider the impact of grid reliability on nuclear plants. The

blackout affected more than 290 commercial power facilities, including ten nuclear plants.

The NRC provided extensive technical support to the joint U.S.-Canada Power System Outage

Task Force to evaluate nuclear power plant responses to the blackout. In April 2004, the NRC

issued a regulatory information summary to inform licensees of nuclear power plants that grid

reliability can impact plant risk and the availability of offsite power.

On February 1, 2006, the NRC issued GL 2006-02, "Grid Reliability and the Impact on Plant Risk

and the Operability of Offsite Power." The purpose of this generic letter was to determine

whether licensees were complying with NRC regulatory requirements governing electric power

sources and associated personnel training. The staff did not identify any safety concerns or

compliance issues as a result of its review of GL 2006-02.

In addition to the efforts described above and because of the importance of the offsite power

system, the NRC continues to monitor grid reliability and works with the organizations

responsible for regulating the grid (i.e., the Federal Energy Regulatory Commission, the U.S.

Department of Energy (DOE), and the Electric Reliability Organization (i.e., the North American

Electric Reliability Corporation)) and licensees to ensure that they are prepared to address

changes in the overall reliability of the offsite power system.

Emergency Preparedness and Security

Emergency Preparedness

After the terrorist attacks on September 11, 2001, the NRC required nuclear facility licensees

to assess the potential impact of a terrorist-initiated event on the site emergency plan. This

additional requirement complements the emergency plan and provides assurance that licensees

are prepared to respond to a terrorist event.

The NRC is continuing to develop emergency preparedness policies, regulations, programs, and

guidelines for both currently licensed nuclear reactors and potential new nuclear reactors;

provide technical expertise on emergency preparedness issues; coordinate, as appropriate, with

other parts of the NRC and stakeholders; and oversee and provide technical direction for the

Emergency Preparedness cornerstone of the Reactor Oversight Process.

Security

The NRC ordered new security-related requirements after September 11, 2001. The NRC is

currently codifying many of the additional requirements imposed on licensees by those orders,

based on experience and comprehensive reviews. For example, the NRC has proposed a

comprehensive revision to the requirements for physical protection at nuclear reactors and

published a proposed rule to amend its regulations pertaining to the design basis threat.

After September 11, 2001, the NRC issued an order (Interim Compensatory Measures Order

EA-02) as part of a comprehensive effort to improve the capabilities of commercial nuclear

20
reactor facilities to respond to terrorist threats. Section B.5.b. of the Order required licensees to

develop specific guidance and strategies to maintain or restore core cooling, containment, and

spent fuel pool cooling capabilities using existing or readily available resources that could be

effectively implemented should the loss of large areas of the plant due to explosions or fire

occur, including those that an aircraft impact might create. Although it was recognized before

September 11, 2001, that nuclear reactors already had significant capabilities to withstand a

broad range of attacks, implementing'these mitigation strategies would significantly enhance the

plants' capabilities to withstand a broad range of threats.

Licensee have been implementing Section B.5.b mitigation strategies since the Order was

issued. In 2005, the NRC issued guidance to more fully describe its expectations for

implementing Section B.5.b. The guidance relied upon lessons learned from detailed NRC

engineering studies and industry best practices. Additionally, the NRC conducted two on-site

team assessments at each reactor facility that identified additional mitigating strategies to

preserve core cooling, containment integrity, and spent fuel pool cooling. The NRC has

incorporated requirements for the B.5.b. mitigating strategies into facilities' operating licenses.

These strategies include fire-fighting response strategies, operations to mitigate fuel damage,

and actions to minimize release of radioactive material. Elements that can be part of these

strategies include an assessment of mutual aid fire fighting assets, developing designated

staging areas for the equipment and materials, communications protocols, training on integrated

fire response strategy, and water spray scrubbing. In total, these enhancements strengthen the

interface between plant safety and security operations and add to the safety approach of

defense-in-depth.

Since September 11, 2001, the issue of an airborne attack on U.S. infrastructure, including both

operating and potential new nuclear power plants, has been widely discussed. The NRC has

comprehensively studied the effect of an airborne attack on nuclear power plants. Studies

confirm that there is a low likelihood that an airplane attack on a nuclear power plant would

affect public health and safety, owing in part to the inherent robustness of the structures. One

study identified new methods plants could use to minimize damage and risk to the public in the

event of any kind of large fire or explosion. Nuclear power plants subsequently implemented

many of these methods. The NRC is now considering new regulations for future reactors'

security regarding inherent safety and security features to minimize potential damage from an

airborne attack.

The third review meeting concluded that one NRC challenge was to ensurethat the

post-September 11, 2001 emphasis on security does not adversely affect operational safety.

One way the NRC is meeting this challenge is by proposing to amend 10 CFR 73, "Physical

Protection of Plants and Materials," to explicitly require licensees to assess and manage the

safety/security interface at commercial power reactors. The proposed amendment will require

(1) licensees to assess and manage the potential for adverse effects between safety and

security (including the site emergency plan) before implementing changes to plant

configurations, facility conditions, or security; (2) changes to be assessed and managed must

, include planned and emergent activities such as physical modifications, procedural changes,

and maintenance activities; (3) that upon identifying potential adverse interactions, licensees

must communicate them to the appropriate licensee personnel and take corrective or

compensatory actions to maintain safety and security in accordance with applicable regulations,

orders, and license conditions. In addition, the NRC has prepared a draft regulatory guide, DG5021, "Managing the Safety/Security Interface," to describe acceptable methods to implement

21
the proposed amendment, explain techniques that the NRC will use in evaluating compliance,

and guide applicants and licensees.

Other Major Regulatory Accomplishments

Since its previous U.S. National Report in 2004, the NRC has issued two early site permits and

amended its regulations concerning the design-basis threat (DBT), the AP1000 design

certification, and risk-informed categorization and treatment of SSCs for nuclear power reactors.

Issuance of Early Site Permits for the Clinton and Grand Gulf Sites

On March 15 and April 5, 2007, the NRC issued thefirst early site permits to the Exelon

Generation Co. for the Clinton site near Clinton, Illinois, and to System Energy Resources

(SERI) for the Grand Gulf site near Port Gibson, Mississippi, respectively. The main advantage

of the early site permit process is the removal of environmental contentions later in the licensing

process. Successful completion of the early site permit process resolves many site-related

safety and environmental issues and determines that the sites are suitable for possible future

construction and operation of a nuclear power plant. The permits are valid for up to 20 years.

An early site permit may be referenced in an application to the NRC for a combined license to

build one or more nuclear plants on the permitted site.

Design-Basis Threat

On March 19, 2007, the NRC amended the regulations at 10 CFR Part 73, "Physical Protection

of Plants and Materials," that govern DBTs. This final rule makes security requirements that are

similar to those previously imposed by the Commission's April 29, 2003 DBT orders, which are

generically applicable to all plants and redefine the level of security requirements necessary to

ensure that the public health and safety, common defense, and security are adequately

protected. The rule revises the DBT requirements for radiological sabotage, generally applicable

to power reactors and Category I fuel cycle facilities, and for theft or diversion of NRC-licensed

strategic special nuclear material, also applicable to Category I fuel cycle facilities.

AP1000 Design Certification

On January 27, 2006, the NRC amended 10 CFR Part 52 to certify Westinghouse's AP1 000

standard plant design. This action was necessary so that applicants intending to construct and

operate an AP1000 design may do so by referencing the regulation (AP1000 Design

Certification Rule).

Risk-Informed Categorization and Treatment of SSCs for Nuclear Power Reactors

On November 22, 2004, the NRC amended 10 CFR Part 50 to provide an alternative approach

for establishing the requirements for treatment of SSCs for nuclear power reactors using a

risk-informed method of categorizing SSCs according to their safety significance. The 10 CFR

50.69 rule revisesrequirements with respect to "special treatment," that is, those requirements

that provide increased assurance (beyond normal industrial practices) that SSCs perform their

design-basis functions. This amendment permits licensees and applicants for licenses to

remove SSCs of low safety significance from the scope of certain, identified special treatment

requirements and revises requirements for SSCs of greater safety significance. In addition to

22
the rulemaking and its associated analyses, the Commission also issued a regulatory guide to

implement the rule.

23

-CONCLUSIONS ON THE U.S. NATIONAL REPORT

FROM THE THIRD REVIEW MEETING

This section presents the conclusions from the review of the 2004 U.S. National Report at the

third review meeting in April 2005.

Delegates from other countries noted that the United States prepared and delivered a highly

informative and candid presentation at the country group meeting. The country group also noted

that the United States provided extensive and comprehensive answers to all of the 266

questions posted by other contracting parties.

Items Resulting from Country Group Session

Review of the questions raised by other contracting parties on the U.S. National Report identified

the following areas of interest:

(1) risk-informed regulation,

(2) licensing and oversight,

(3) long-term operation,

(4) safety culture and human factors,

(5) operating experience,

(6) organizational structure,

(7) emergency preparedness and security, and

(8) radiation protection.

The NRC's presentation to Country Group 1 focused on these topics.

Country Group I participants concluded that the U.S. good practices were:

(1) the transition toward risk-informed approaches in an incentive-based manner,

(2) the successful implementation of the Reactor Oversight Process,

(3) the NRC's openness as demonstrated by its Strategic Plan, improved Web site, and

commitment to stakeholder involvement, and

(4) the NRC's comprehensive and innovative strategies to attract and retain staff.

Among NRC's challenges, Country Group I identified:

(1) ensuring that the post-September 11, 2001 emphasis on security does not have an

adverse effect on operational safety,

(2) continuing the focus on safe long-term operation and emerging technical issues,

(3) preparing for the potential licensing of new reactors,

(4) appropriately emphasizing safety culture, safety management, and organizational

aspects of nuclear safety,

(5) continued monitoring and analysis of operating experience and timely implementation of

lessons learned, and

(6) expanding risk-informed, performance-based concepts while ensuring a consistent

regulatory approach..

25
The group's highlighted planned U.S. initiatives included:

(1) enhancing the Reactor Oversight Process to increase NRC inspector review of licensee

safety-conscious work environment,

(2) working with the international community to define the ways and means to detect

deteriorating performance, including safety culture and safety management, and

(3) planning for the performance of an International Regulatory Review Team (IRRT)

self-assessment (since the time of the previous report, this group has been renamed the

Integrated Regulatory Review Service (IRRS)).

The current U.S. National Report addresses many of these issues under the relevant articles.

Specifically, Article 10 discusses risk-informed regulation, Article 6 presents the Reactor

Oversight Process, Article 14 addresses license renewal and periodic safety reviews, and

Articles 17 and 18 cover new reactor licensing. Article 8 discusses the IRRS.

NRC Areas for Improvement and Major Challenges for the Future

The NRC identified major challenges for the future in its Strategic Plan 2004-2009, which-were

unchanged from those identified in the third U.S. National Report. Challenges arise from the %

changing regulatory environment and external factors. The challenges are summarized below;

Appendix A to this report provides more details.

The NRC will strengthen the interrelationship among safety, security, and emergency

preparedness.

The majority of operating nuclear power plants will have applied for license renewal to

help meet the country's demand for energy. A challenge is to monitor, manage, and

control the effects of aging so that safety is ensured for the renewal period.

The U.S. Department of Energy (DOE) will apply to construct and operate the country's

high-level radioactive waste repository. The timing of this action, may challenge the

allocation of the NRC's resources.

The U.S. nuclear power industry will show a growing interest in licensing and

constructing new nuclear power plants to meet the Nation's demand for energy.

Challenges include analyzing in detail the vulnerability to accidents and security

compromises, as well as developing appropriate inspections, tests, analyses, and

acceptance criteria for construction.

The NRC will continue to see increased requirements to coordinate with a wide array of

Federal, State, and local agencies related to homeland security and emergency planning.

* The regulatory climate is expected to adjust to both internal and external factors

(described below).

Key external factors that could cause challenges are:

* Receipt of new reactor operating license applications

26
a A significant operating incident (domestic or international)

* A significant terrorist incident

Timing of the DOE application and related activities for the high-level waste repository at

Yucca Mountain

Homeland security initiatives

Legislative initiatives

NRC Maior Management Challenges for the Future

By law, the Inspector General of each Federal agency (discussed in Article 8) identifies the

agency's most serious management and performance challenges facing the agency and

assesses progress in addressing those challenges. The NRC's Inspector General's annual

assessment of the major management challenges confronting the agency can be found on

NRC's public Web site., Eight of the nine challenges are essentially the same as those identified

in the previous U.S. National Report. In 2006, the Inspector General identified a new challenge

titled, Ability to Meet the Demand for Licensing New Reactors, and removed a challenge titled,

Intraagency Communication. The main challenges are summarized below. For more detail, see

Appendix B to this report.

Protection of nuclear material used for civilian purposes

Protection of information

Development and implementation of a risk-informed and performance-basedregulatory

oversight approach

Ability to modify regulatory processes to meet a changing environment

Implementation of information resources

Administration of all aspects of financial management

Communication with external stakeholders throughout the NRC's regulatory activities

Managing human capital

Ability to meet the demand for licensing new reactors

27

PART 2

29

ARTICLE 6. EXISTING NUCLEAR INSTALLATIONS

Each Contracting Party shall take the appropriate steps to ensure that the safety

of nuclear installations existing at the time the Convention enters into force for that

Contracting Party is reviewed as soon as possible. When necessary in the context of this

Convention, the Contracting Party shall ensure that all reasonable practicable

improvements are made as a matter of urgency to upgrade the safety of the nuclear

installation. If such upgrading cannot be achieved, plans should be implemented to shut

down the nuclear installation as soon as practically possible. The timing of the shutdown

may take into account the whole energy context and possible alternatives, as well as the

social, environmental, and economic impact.

This section explains how the United States ensures the safety of nuclear installations

in accordance with the obligations in Article 6. This section covers the reactor licensing and

major oversight processes in the United States. This section also discusses programs for

rulemaking, fire protection regulation, decommissioning, research, and programs for public

participation. The NRC posts the major results of assessments on the agency's public Web site

at http://www.nrc.qov. This update includes expectations about early site permits and design

certification applications, current experience, and revised details about programs.

6.1 Introduction

The NRC's primary goal is safety. The Agency achieves this goal by ensuring that the

performance of licensees is at or above acceptable safety levels. The NRC's licensees are

responsible for designing, constructing, and operating nuclear facilities safely, while the NRC is

responsible for the regulatory oversight of the licensees. Five strategic outcomes for this goal

are specified:

• No nuclear reactor accidents.

- No inadvertent criticality events.

- No acute radiation exposures resulting in fatalities.

- No releases of radioactive materials that result in significant radiation exposures.

- No releases of radioactive materials that cause significant, adverse environmental

impacts.

The NRC met all of its safety strategic outcomes in fiscal years (FYs) 2005 and 2006.

The NRC also uses performance measures to determine whether the Agency has met its safety

goal. The NRC met its performance measures in FYs 2005 and 2006. Currently the NRC uses

six performance measures.

The first measure analyzes plant performance based on alarge number of performance

indicators and inspection findings.

The second measure tracks significant precursor events determined by the likelihood of an event

adversely impacting safety.

31
The third performance measure indicates whether the NRC identifies significant issues in a plant

during inspections conducted under the reactor oversight program.

The fourth measure tracks the trends of several key indicators of nuclear power plant safety.

This measure is the broadest measure of the safety of nuclear power plants, incorporating the

performance results from all plants to determine industry average results.

These measures indicated that not only were the plants safely operated, but the events that did

occur were of relatively minor significance.

The other two measures address harmful radiation exposures to the- public and occupational

workers and radiation exposures that harm the environment. Neither of these measures

exceeded their targets in FY 2006.

6.2 Nuclear Installations in the United States

For a list of all 104 nuclear installations in the United States, see Annex 1. The source of this list

is the NRC Information Digest, 2006-2007 edition, available on the agency's Web site.

6.3 Regulatory Processes and Proarams

6.3.1 Reactor Licensing

To construct and operate a nuclear reactor, an entity must submit an application to the NRC for

a safety and environmental review. The public has opportunities to participate through a hearing

process. The NRC licensed all current operating nuclear plants under the detailed. two-step

process specified in 10 CFR Part 50, first issuing a construction permit and then an operating

license. Although the NRC has not received any applications for reactor licenses since 1976, it

has recently received applications for early site permits and design certifications. In addition,

due largely to the favorable incentives created by the U.S. Congress in the Energy Policy Act of

2005, the industry has indicated that it will be submitting applications for two additional design

certifications, and potentially as many as 20 applications for up to 28 reactor licenses during

2007 through 2009. The agency will review these applications under a new licensing process,

specified in 10 CFR Part 52. Article 18 provides more detail about the 10 CFR Part 52

regulations.

The NRC's reactor licensing process provides for the review and approval of changes after initial

licensing. The process allows amendments to the operating license to support plant changes,

license renewal, changes of ownership and license transfer, exemptions and relief from NRC

regulations, and increasing the reactor power level (i.e., power uprates). The process is

discussed further in the Introduction and in other articles (Articles 14, 17 and 18).

6.3.2 Reactor Oversight Process

Through its Reactor Oversight Process the NRC continuously oversees nuclear power plants to

verify that they are being operated in accordance with the agency's rules and regulations. NRC

has full authority to take whatever action is necessary to protect public health and safety, and

may demand immediate licensee actions, up to and including a plant shutdown.

32
The Reactor Oversight Process uses both inspection findings and performance indicators to

assess the performance of each plant within a regulatory framework of seven cornerstones of

safety. Toward that end, the NRC performs a program of baseline inspections at each plant and

may perform supplemental inspections and take additional actions to ensure that the plants

address significant issues. The NRC communicates the results of its oversight process by

posting plant-specific inspection findings and performance indicator information on the NRC's

public Web site. The NRC also conducts public meetings with licensees to discuss the results of

the NRC's assessments of licensee performance.

NRC assesses the Reactor Oversight Process annually and evaluates the overall effectiveness

of the Reactor Oversight Process through its success in meeting its preestablished goals (i.e.,

performance metrics) and intended outcomes. The latest report, SECY-07-0069, "Calendar

Year 2006 Reactor Oversight Process Self-Assessment," was issued on April 6, 2007.

The calendar year 2006 self-assessment results indicated that the Reactor Oversight Process

was successful in meeting its program goals and achieving its intended outcomes. The Reactor

Oversight Process was deemed to be objective, risk-informed, understandable, and predictable,

and met the agency goals of ensuring safety, openness, and effectiveness. The NRC staff

maintained its focus on stakeholder involvement and continued to improve various aspects of

the Reactor Oversight Process as a result of feedback and lessons learned. The NRC staff

implemented several process improvements in 2006 to address issues raised by the

Commission, recommended, by independent reviews, and obtained from internal and external

stakeholder feedback. Most notably, the NRC staff made significant enhancements to the

assessment and inspection programs in 2006 to more fully address safety culture.

The NRC continued to improve the performance indicator program to better identify declining

plant performance in a timely manner, but recognized the need for further improvement. The

inspection program independently verified that plants were operated safely, appropriately

identified performance issues, and ensured the adequacy of licensee corrective actions to

address the noted performance issues. Further improvements in the significance

determination process resulted in the timeliness goal for making final determinations being met

in 2006 for the first time since implementation. [The significance determination process uses

risk insights to help NRC inspectors and staff determine the safety significance of inspection

findings.] The assessment program ensured the NRC and licensees took necessary actions to

address identified performance issues. The NRC continues to actively solicit input from the

NRC's internal and external stakeholders and expects further process improvements based on

stakeholder feedback and lessons learned.

Based on its calendar year 2006 self-assessment, the NRC intends to focus on the following

significant actions or activities to improve the efficiency and effectiveness of the Reactor

Oversight Process in 2007: (1) continue to monitor implementation of the Mitigating Systems

Performance Index and address additional improvements to the performance indicator program

to better identify those plants with declining performance; (2) implement the Reactor Oversight

Process realignment process and adjust inspection resources accordingly; (3) continue to

monitor implementation of the safety culture enhancements and address related

recommendations in this area; and (4) implement adjustments to the process to incorporate

pending Commission direction related to the point at which licensee senior management will be

requested to meet with the Commission to discuss actions being taken to improve performance.

33
1 6.3.3 Industry Trends Program

The NRC staff implemented the Industry Trends Program in 2001 and is continuing to develop

the program as a means to confirm that the nuclear power industry is maintaining the safety of

operating power plants and to increase public confidence in the effectiveness of the NRC's

processes. The agency uses industry-level indicators to identify adverse trends in performance.

After assessing industry trends for safety significance, the NRC responds as necessary to any

identified safety issues, including adjusting the inspection and licensing programs if necessary.

One important output of the Industry Trends Program is the annual agency performance

measures reported to Congress on the number of statistically significant adverse industry trends.

The NRC Performance and Accountability Report includes this outcome measure.

In addition to long-term trending of the data to identify statistically significant adverse trends, the

NRC staff uses a statistical approach based on prediction limits to identify potential short-term,

year-to-year emergent issues before they become long-term trends. Short-term trending of the

fiscal year 2006 data did not identify any issues that warranted additional analysis or significant

adjustments to the nuclear reactor safety inspection or licensing programs.

The Reactor Oversight Process uses both plant-level performance indicators and inspections

to provide plant-specific oversight of safety performance, whereas the Industry Trends Program

provides a means to assess overall industry performance using industry-level indicators. The

NRC evaluates issues that are identified from either program using information from agency

databases and addresses those determined to have generic safety significance, including

generic safety inspections under the Reactor Oversight Process, the generic communications

process, and the generic safety issue process.

Based on the information currently available from the industry-level indicators and the Accident

Sequence Precursor Program (below), no statistically significant adverse industry trends have

been identified through FY 2006. For more details, see SECY-07-0063, "FY 2006 Results of the

Industry Trends Program for Operating Power Reactors-and Status of Ongoing Development,"

dated April 3, 2007, which is available on the NRC public Web site.

6.3.4 Accident Sequence Precursor Program

The Accident Sequence Precursor Program systematically evaluates U.S. nuclear power plant

operating experience to identify, document, andi rank the operating events that are most likely to

have led to inadequate core cooling and severe core damage (precursors), accounting for the

likelihood of additional failures. The objectives of the program include the following:

(1) provide a comprehensive, risk-informed view of nuclear power plant operating experience

and a measure for trending nuclear power plant core damage risk

(2) provide a partial check on dominant core damage scenarios predicted by PRAs

(3) provide feedback for regulatory activities

To identify potential precursors, the NRC reviews plant events from licensee event reports, NRC

inspection reports, and special requests from NRC staff. The NRC then analyzes any identified

potential precursor by calculating a probability of an event leading to a core damage state. The

34
program considers a precursor to be an event with a conditional core damage probability

(CCDP) greater than or equal to 1 x10-6 and a significant precursor as an event with a CCDP

greater than or equal to 1 x10-3.

In 2006, the NRC streamlined the analysis and review processes to improve the timeliness and

reduce the number of Accident Sequence Precursor Program analyses that undergo formal

peer review. The analysis process now includes results from the significance determination

process and the NRC Incident Investigation Program when practicable.

The staff completed precursor trend analyses as part of the annual Accident Sequence

Precursor Program status report provided to the Commission in SECY-06-0208, "Status of the

Accident Sequence Precursor Program and the Development of Standardized Plant Analysis

Risk Models," dated October 5, 2006. The following are a few of the insights provided in the

report.

(1) The staff identified no significant precursors in FY 2005 and F Y 2006. In addition, the

staff identified no statistically significant trendsfor significant precursors during the FY

1996-2005 period.

(2) The risk contribution from precursors was generally constant during the FY 1996-2003

period and has decreased during FY 2003-2005.

(3) The staff evaluated precursor data through FY 2005 to identify statistically significant

adverse trends for the Industry Trends Program. The staff divided the data for the rates

of occurrence of all precursors during the FY 1996-2005 period into two statistically

distinct groups, FY 1996-2000 and FY 2001-2005, because of program scope changes

that occurred in FY 2000. The analysis detected no statistically significant trend for

either of these 5-year periods.

(4) Over the more recent time period from FY 2001-2005, the analysis detected a

statistically significant decreasing trend for higher risk precursors (i.e., CCDP 104).

6.3.5 Operating Experience Program

The NRC launched a revised Operating Experience Program in January 2005, recognizing that

the effective use of operating experience is important for the safety mission of the agency.

Under the current NRC Strategic Plan, the agency is committed to "evaluate and utilize domestic

and international operational experience and events to enharnce decision-making," as part of its

effort to achieve the goal of safety. As a result, the NRC's emphasis on the effective use of

operating experience remains strong.

The fundamental aim of the Operating Experience Program is to collect, evaluate, communicate,

-and apply operating experience information to achieve the NRC's principal safety mission of

protecting people and the environment. Operating experience is reported to the NRC or

identified in licensee event notifications, and in many other reports that are submitted under

licensee reporting requirements of the regulations, and in reports of operating experience at

foreign facilities. Sources of foreign operating experience include International Nuclear Event

Scale (INES) events and Incident Reporting System reports. NRC staff systematically screens

nuclear rea6tor related operating experience for safety significance and generic implications.

35
The NRC staff also determines the need for further action and application of lessons learned

related to plant operating experience.

To support its safety mission, the NRC increased resources dedicated to operating experience

review and instituted a clearinghouse. The clearinghouse collects, stores, screens, and

communicates operating experience; conducts and coordinates the evaluation of operating

experience; tracks the application of operating experience lessons learned; and coordinates the

NRC operating experience activities with other organizations performing related functions.

Upon launching the program, the NRC developed an internal Web site to provide a centralized

source for accessing reactor operating experience information. This Web site is a gateway to

the agency's operating experience document collections, contacts, search tools, sources, and

reference material. In addition, the NRC created an operating experience community forum to

quickly disseminate operating experience to the appropriate technical staff. All. of the NRC's

event-related reports can be found on the agency's public Web site.

For more information on this program, see Section 19.7 of this report.

6.3.6 Program for Resolving Generic Issues

The program for resolving generic issues provides an effective means for addressing issues that

are not sufficiently addressed by existing rules, guidance, or programs and that affect licensees

(and others under NRC jurisdiction). The NRC maintains a complete list of all generic issues in

NUREG-0933, "A Prioritization of Generic Safety Issues." The most recent revision, published in

October 2006, is available on the agency's public Web site.

Sources of candidate generic issues include safety evaluations, operational events, and

suggestions from individual staff members, outside organizations, or members of the general

public. Existing programs generally address emergent issues that demand immediate attention

(e.g., issues that may require plant shutdown) so that quick decisions can be made.

6.3.7 Rulemaking

The NRC's regulations, also called rules, impose requirements that licensees must meet to

obtain or retain a license or certificate to use nuclear materials or to operate a nuclear facility.

The technical staff usually proposes a rule or a change to a rule because of a perceived need to

protect the public health and safety. However, any member of the public may petition the NRC

to develop, change, or rescind a rule. The impetus for a proposed rule could be a requirement

issued by the Commission, a petition for rulemaking submitted by a member of the public, or

research results that indicate a need for a rule change. The NRC publishes the proposed rule in

the Federal Register for public comment. Once the public comment period has closed, the staff

analyzes the comments, makes any needed changes, and forwards the final rule for approval,

signature, and publication in the Federal Register.

The NRC developed RuleForum to provide an easy means for members of the public to access

and comment on NRC rulemaking actions. Accessible through NRC's public Web site,

RuleForum contains proposed rulemakings that have been published in the Federal Register,

petitions for rulemaking, and other types of documents related to rulemaking proceedings.

36
The NRC Commissioners must approve each final rule that involves significant matters of policy.

Once approved, the final rule is published in the Federal Register and usually becomes effective

30 days later. The introduction to this report summarizes the significant nuclear reactor-related

rules issued since the previous U.S. National Report.

6.3.8 Fire Regulation Program

The NRC has three main foci in fire protection regulation. These foci are: implementation of the

new risk-informed, performance-based fire protection licensing basis (10 CFR 50.48(c));

resolution of the fire-induced multiple spurious operation/circuit analysis issue; and resolution of

licensees' non-conforming post-fire operator manual actions. To support the implementation of

10 CFR 50.48(c), the NRC issued Regulatory Guide 1.205 "Risk-Informed, Performance-Based

Fire Protection for Existing Light-Water Nuclear Power Plants," in May 2006. As of September

2007, there are 42 reactor units committed to transitioning to 10 CFR 50.48(c). Two nuclear

stations, Oconee and Shearon Harris, volunteered to be pilot plants for the transition, and the

pilots' license amendment requests are expected in May and June 2008. Regarding circuit

analysis actuations, the NRC continues to engage with stakeholders to develop an acceptable

method to resolve the issue. On the topic of operator manual actions, the NRC has issued

Regulatory Issue Summary 2006-10 "Regulatory Expectations with Appendix R Paragraph

III.G.2 Operator Manual Actions," and developed NUREG-1852 "Demonstrating the Feasibility

and Reliability of Operator Manual Actions in Response to Fire." Licensees are to resolve issues

involving post-fire operator manual actions by March 2009.

The NRC has an active fire research program that develops the technical bases for ongoing and

future regulatory activities in fire protection and fire risk analysis. NRC currently has ongoing

research in (1) developing and improving fire risk analysis methods and tools; (2) applying these

methods and tools to develop risk insights; (3) collecting, generating and analyzing fire related

data; (5) verifying, validating and improving fire models for regulatory use; (6) performing

specialized fire testing on electrical cables for both hot shorts and fire properties; (7) evaluating

shipping casks for beyond design basis fire conditions; and (8) evaluating methods to predict

operator performance during fire conditions.

The fire research program supports the Agency's strategic goals of safety and effectiveness and

partners with other organizations with similar missions such the National Institute of Standards

and Technology, the Electric Power Research Institute (EPRI) and on the international level with

groups such as Organization for Economic and Cooperation and Development Committee on

the Safety of Nuclear Installations.

6.3.9 Decommissioning

The decommissioning process consists of a series of integrated activities, beginning with the

nuclear facility transitioning from "active" to "decommissioning" status and concluding with

termination of the license, and release of the site. The NRC has adopted extensive regulations

to ensure that decommissioning is accomplished safely and that residual radioactivity is reduced

to a level that permits release of the property for unrestricted use. The NRC reviews and

approves license termination plans, conducts inspections, processes license amendments, and

monitors the status of activities to ensure that radioactive contamination is reduced or stabilized.

37
The design criteria for new facility construction at 10 CFR 20.1406, "Minimization of

Contamination," address early consideration of decommissioning: Furthermore, the safety

standards on decommissioning promulgated by the International Atomic Energy Agency (IAEA)

cite considerations, which the United States supports, for future decommissioning provisions in

the conceptual design of nuclear facilities.

Spent fuel can remain stored in the spent fuel pool or in dry cask storage facilities until a

geologic repository is built and operating. The NRC regulations in 10 CFR Part 50 and

10 CFR Part 72, "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel,

High-Level Radioactive Waste, Reactor-Related Greater Than Class C Waste," contain licensing

requirements to maintain spent fuel integrity. The Commission, in issuing its Waste Confidence

Decision in 1990, found that spent fuel can be stored safely in spent fuel pools or in onsite

independent spent fuel storage installations without significant environmental impacts for at least

30 years beyond the plant's licensed life (which may include the term of a renewed license).

6.3.10 Reactor Safety Research Program

The NRC conducts reactor safety research to support its mission of ensuring that its licensees

safely design, construct, and operate nuclear reactor facilities. The agency carries out this

research program to identify, evaluate, and resolve safety issues; to ensure that an independent

technical basis exists to review licensee submittals; to evaluate operating experience and results

of risk assessments for safety implications; and to support the development and use of riskinformed regulatory approaches. In conducting the Reactor Safety Research Program, the NRC

anticipates challenges posed by the introduction of new technologies. The NRC continues to

seek out opportunities to leverage its resources through both domestic and international

cooperative programs and to provide enhanced opportunities for stakeholder involvement and

feedback on its research program.

The NRC conducts pre-application reviews for advanced non-light water reactor designs under

the safety research program. In the pre-application phase, NRC interacts with prospective

design certification applicants to address topics that would benefit both the applicant and the

staff in preparing for a design certification application. The Commission's Policy Statement on

Advanced Reactors (SECY-93-087, "Policy, Technical, and Licensing Issues Pertaining to

Evolutionary and Advanced Light-Water Reactor Designs,") encourages early interactions on

such advanced designs so as to facilitate the resolution of safety issues early in the design

process. In addition, the agency will conduct research to address technical issues that it

anticipates will arise during its review of advanced reactor designs.

6.3.11 Special Programs for Public Participation

The NRC values public participation in its regulatory processes.. To this end, the NRC provides

the diverse body of stakeholders (general public; Congress; other Federal, State, and local

governments; Indian Tribes; industry; technical societies; the international community; and

citizen groups) clear and accurate information about its role and opportunities to participate in

the agency's regulatory programs. Numerous NRC programs and processes provide the public

with accessibility to NRC staff andresources, seek to make communication with stakeholders

more clear, accurate, reliable, objective, and timely, and help to ensure that the reporting of the

performance of nuclear power plants performance is open and objective. The agency has

developed Web pages to disseminate timely, accurate information regarding issues of interest to

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the public or events at nuclear facilities. The NRC elicits public involvement early in the

regulatory process so as to address any safety concerns in a timely manner. In addition to the

formal petition and hearinig processes integrated into the licensing program, the agency also

uses feedback forms at public meetings to obtain public input.

Fostering an environment in which safety issues can be openly identified without fear of

retribution is of paramount interest to the NRC. The agency has established tools for the public,

industry, and NRC employees to use to raise safety concerns, including the petition process

under 10 CFR 2.206, "Requests for Actions under this Subpart," the safety-conscious work

environment policy, and the allegation program.

The NRC petition process regulations in 10 CFR 2.206 allow any member of the public to raise

potential health and safety concerns and ask the agency to take specific enforcement actions

against a licensee. If warranted, the NRC can modify, suspend, or revoke a license, or take

other appropriate enforcement action, to resolve a problem identified in the petition. Recent

changes made to the petition process emphasize a timely response to the petitioner and

encourage increased, direct involvement of the petitioner (in addition to involvement of the

licensee) by allowing the petitioner to personally address the petition review board and comment

on the agency's decision.

Any member of the public may petition the NRC to develop, change, or rescind a rule under 10

CFR 2.802. Upon receiving the petition, NRC publishes it in the Federal Register for public

comment. The NRC staff will evaluate the petition and any comments received and may either

grant or deny the petition or, in some instances, may partially grant/ deny the petition. If granting

the petition, the NRC will publish a proposed rule that would address the concern included in the

petition. This action would be followed by the publication of a final rule. If denying a petition,

NRC publishes a notice of denial in the Federal Register. This notice of denial will address any

public comments received and the reason for denying the petition.

NRC encourages workers in the nuclear industry to take their concerns directly to their

employers and is particularly vigilant about a safety-conscious work environment that

encourages such reporting. It is NRC's expectation that licensees establish and maintain a work

environment where employees do not fear retribution by a licensee for raising concerns about

safety or regulatory issues. Additionally, workers and members of the public may bring their

concerns relating to safety or regulatory issues directly to the NRC. The agency established a

toll-free safety hotline for reporting such concerns, and NRC management, staff, and inspectors,

including the resident inspectors at plant sites, are trained and available to receive such

concerns.

Historically, approximately 500 potential allegations have been reported directly to the NRC

allegation program each year by industry workers or members of the public. The NRC

developed the allegation program to establish a formal process for evaluating and responding to

each issue. The primary purpose of the program is to provide an alternative method for

individuals to raise safety or regulatory issues and to have them addressed. About 60 percent of

the issues that are reported to the NRC are from licensee employees, employees of contractors

to licensees, or former employees of licensees or contractors. Given sufficient information, the

staff will evaluate each issue to determine whether it can verify the issue and, if so, the effect of

the issue on plant safety; The evaluation either involves an engineering review, inspection, or

investigation by NRC staff or an evaluation by the licensee that is assessed by the NRC staff.

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-Historically, the NRC has been able to substantiate 25 to 30 percent of the allegations received.

If the evaluation reveals a violation of regulatory requirements, the agency takes appropriate

enforcement action. Additionally, the NRC informs in writing the individual who raised the issue

of the results of its evaluation, except in limited instances when sensitive security related matters

are discussed.

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ARTICLE 7. LEGISLATIVE AND REGULATORY FRAMEWORK

1. Each Contracting Party shall establish and maintain a legislative and regulatory

framework to govern the safety of nuclear installations.

2. The legislative and regulatory framework shall provide for:

(i) the establishment of applicable national safety requirements and

regulations

(ii) a system of licensing with regard to nuclear installations and the

prohibition of the operation of a nuclear installation without a license

(iii) a system of regulatory inspection and assessment of nuclear installations

to ascertain compliance with applicable regulations and the terms of

licences

(iv) the enforcement of applicable regulations and of the terms of licences,

including suspension, modification, and revocation

This section explains the legislative and regulatory framework governing the U.S. nuclear

industry. It discusses the provisions of that framework for establishing national safety

requirements and regulations and systems for licensing, inspection, and enforcement.

The framework and provisions have not changed since the previous U.S. National Report

was issued. This update includes a revised discussion of 10 CFR Part 52.

7.1 Legislative and Regulatory Framework

The Atomic Energy Act, passed by Congress and signed by the President, provided

the framework for all subsequent regulation of nuclear installations. However, as is generally the

case with most laws, this act provided general principles and concepts and left the regulatory

body (i.e., the NRC) to address the details through specific regulations.

7.2 Provisions of the Legislative and Regulatory Framework

7.2.1 National Safety Requirements and Regulations

The regulations on nuclear installations in the United States are governed by the Atomic Energy

Act and the Energy Reorganization Act. (The Energy Reorganization Act abolished the U.S.

Atomic Energy Commission, and created the NRC and the U.S. Energy Research and

Development Administration (ERDA). ERDA was subsequently subsumed by DOE. The NRC

administers these statutes by licensing commercial nuclear installations in the United States. In

addition, several statutes (listed in previous U.S. National Reports) have substantial bearing on

the practices and procedures of the Commission.

The NRC must license all commercial nuclear installations in the United States. (Some

Government facilities that are operated by or for DOE are exempt from licensing under the

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Atomic Energy Act and the Energy Reorganization Act.) The licensing process is similar in

many respects for reactors, separation facilities, reprocessing plants, and nuclear waste storage

and disposal facilities. The following discussion describes the licensing practices for nuclear

power plants.

7.2.2 Licensing of Nuclear Installations

Chapter 10, Section 103, of the Atomic Energy Act grants the NRC authority to.issue licenses for

nuclear reactor facilities. In addition, Section 103 states that such licenses are subject to such

conditions as the NRC may by rule or regulation establish to effectuate the purposes and

provisions of the Atomic Energy Act.

Section 189a. of the Atomic Energy Act provides affected parties with hearing rights in

proceedings for the granting, suspending, revoking, or amending of a license or construction

permit. Hearings, which are used in licensing proceedings for production and utilization facilities

(e.g., nuclear power plants), are held under procedural rules stated in 10 CFR Part 2, "Rules of

Practice for Domestic Licensing Proceedings and Issuance of Orders," and, in particular,

Subpart C, "Rules of General Applicability." The staff participates as a party in most formal

hearings and may also participate as a party in less formal hearings. Hearings are usually held

before a three-member Atomic Safety and Licensing Board, which is generally composed of one

lawyer and two technical members.

Article 18 describes the licensing process in greater detail. Two alternative approaches to

licensing exist. The traditional approach, under 10 CFR Part 50, requires two steps. In the first

step, the NRC reviews a preliminary application and decides whether to grant a construction

permit. In the second step, the agency reviews the final -application and grants an operating

license. The NRC licensed all current operating plants in the United States according to this

process.

In 1989, the Commission established an alternative licensing system, published in

10 CFR Part 52, which provides for certified standard designs and combined licenses that

resolve design issues before construction, and early site permits that resolve most siting issues

years before construction. The basic concept underlying 10 CFR Part 52 is that nuclear reactor

designs can be approved through generic rulemaking. Once the designs are approved, an

applicant can reference them in applications for permission to build and operate nuclear power

plants without the necessity to relitigate, in individual hearings, the issues resolved in the

rulemaking. Moreover, the NRC will determine and approve before construction the criteria for

evaluating whether the plant had been built as specified. Thus, the plant could begin operation

without a second hearing, provided that it satisfied the acceptance criteria. To the extent

possible, issues would be litigated before construction, not once construction is nearly complete,

when the consequences of delay are much greater. In adopting 10 CFR Part 52, the

Commission used the latitude allowed by law to streamline licensing.

Recently, NRC amended Part 52 to improve the effectiveness of its processes for licensing

future nuclear power plants. The amendments clarify the overall regulatory relationship between

Part 50 and Part 52, reorganize Part 52, and reconcile differences in wording in other parts of

the regulations to provide consistent terminology throughout all of the regulations affecting part

52. The amendments also added new sections on written communications, employee

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protection, completeness andaccuracy of information, exemptions, combining licenses, and

jurisdictional limits.

7.2.3 Inspection and Assessment

Under the Atomic Energy Act, the NRC has the authority to inspect nuclear power*plants in its

role of protecting public health and safety and the common defense and security. The staff

inspects power reactors under construction, in test conditions, and in operation to ascertain

compliance with regulations and license conditions. Through its inspection program, the NRC

assesses whether activities are properly conducted and equipment is properly maintained to

ensure safe operations. The agency integrates inspection results into its overall evaluation of

licensee performance, as discussed in Article 6. If a safety problem exists, or there is a failure to

comply with requirements, the licensee must take prompt corrective action. If necessary, the

NRC may take enforcement action.

7.2.4 Enforcement

The NRC's enforcement jurisdiction is drawn from the Atomic Energy Act and the Energy

Reorganization Act.

Section 161 of the Atomic Energy Act authorizes the NRC to conduct inspections and

investigations and to issue orders as may be necessary or desirable to promote the common

defense and security, protect health, or minimize danger to life or property. Section 186

authorizes the NRC to revoke licenses under certain circumstances (e.g., for material false

statements, for a change in conditions that would have warranted NRC refusal to grant a license

on an original application, for a licensee's failure to build or operate a facility in accordance with

the terms of the permit or license, and for violation of an NRC regulation). Section 234

authorizes the NRC to impose. monetary civil penalties not to exceed $100,000 per violation per

day; however, that amount is adjusted every 4 years by the Federal Civil Penalties Inflation

Adjustment Act and is currently $130,000. In addition to the provisions mentioned in Section

234, Sections 84 and 147 authorize the imposition of civil penalties for violations of regulations

implementing those provisions. Section 232 authorizes the NRC to seek injunctive or other

equitable relief for Violation of regulatory requirements.

Section 206 of the Energy Reorganization Act authorizes the NRC to impose civil penalties on

licensees for knowing and conscious failures to provide the agency with certain safety

information.

Chapter 18 of the Atomic Energy Act provides for varying levels of criminal penalties (i.e.,

monetary fines and imprisonment) for willful violations of the act or the regulations or orders

issued under Sections 65, 161b) 161i, or 161o of the act.

Section 223 allows criminal penalties to be imposed on certain individuals who are employed by

firms constructing or supplying basic components of any utilization facility if the individual

knowingly and willfully violates NRC requirements in a manner that could significantly impair a

basic component. Section 235 allows criminal penalties to be imposed on persons who interfere

with nuclear inspectors. Section 236 allows criminal penalties to be imposed on persons who

attempt to or cause sabotage at a nuclear facility or to nuclear fuel. The agency refers alleged

43
or suspected instances of criminal violations of the Atomic Energy Act to the U.S. Department of

Justice for appropriate action.

Subpart B, "Procedure for Imposing Requirements by Order, or for Modification, Suspension, or

Revocation of a License, or for Imposing Civil Penalties," of 10 CFR Part 2 specifies the

procedures that the NRC uses in exercising its enforcement authority. The scope of Subpart B

includes the procedures described below:

(1) 10 CER 2.201, "Notice of Violation," outlines the procedure for issuing notices of

violations.

(2) 10 CFR 2.202, "Orders," explains the procedure for issuing orders. In accordance with

this section, the NRC may decide to issue an order to institute a proceeding to modify,

suspend, or revoke a license or to take other action against a licensee or other person

subject to the NRC's jurisdiction. The licensee or any other person adversely affected by

the order may request a hearing. The NRC is authorized to make orders immediately

effective if required to protect public health, safety, or interest, or if the violation is willful.

(3) 10 CFR 2.204, "Demand for Information," specifies the procedure for issuing a demand

for information to a licensee or other person subject to the Commission's jurisdiction to

.determine whether an order should be issued or other enforcement action should be

taken. The demand does not provide hearing rights because only information is being

sought. A licensee must answer a demand. An unlicensed person may answer a

demand either by providing the requested information or by explaining why the demand

should not have been issued.

(4) 10 CFR 2.205, "Civil Penalties," describes the procedure for assessing civil penalties.

The NRC initiates the civil penalty process by issuing a notice of violation and proposed

imposition of a civil penalty. The agency provides the person charged with an

opportunity to contest in writing the proposed imposition of a civil penalty. After

evaluating the response, the NRC may mitigate, remit, or impose the civil penalty. If the

agency imposes a civil penalty, it provides an opportunity for a hearing. If a civil penalty

is not paid following a hearing, or if a hearing is not requested, the agency may refer the

matter to the U.S. Department of Justice to institute a civil action in Federal District Court

to collect the penalty.

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ARTICLE 8. REGULATORY BODY

1. Each Contracting Party shall establish or designate a regulatory body entrusted

with the implementation of the legislative and regulatory framework referred to in

Article 7, and provided with adequate authority, competence, and financial and

human resources to fulfill its assigned responsibilities.

2. Each Contracting Party shall take the appropriate steps to ensure an effective

separation between the functions of the regulatory body and those of any other

body or organization concerned with the promotion or utilization of nuclear

energy.

This section explains the establishment of the U.S. regulatory body (i.e., the NRC). It also

explains how the functions of the NRC are separate from those of bodies responsible for

promoting and using nuclear energy (e.g., DOE). This update was reorganized. It reports on

the establishment of the NRC's Office of New Reactors and Office of Federal and State

Materials and Environmental Management Programs; organizational changes; current activities;

budget and workforce planning; and the IRRS assessment.

8.1 The Regulatory Body

This section explains the NRC's mandate, authority and responsibilities, structure, international

responsibilities and activities, financial and human resources, position in the governmental

structure, and report of the IRRS self-assessment team.

8.1.1 Mandate

As discussed in Article 7, Congress created the NRC as an independent regulatory agency in

January 1975, with the passage of the Energy Reorganization Act. In giving the NRC an

exclusively regulatory mandate, the statute reflected (in part) a Congressional judgment that the

expanding commercial nuclear power industry (which was expected to continue to grow)

warranted the full-time attention of an exclusively regulatory agency. In creating the NRC,

Congress also addressed a developing public concern that regulatory responsibilities were

overshadowed by the promotion of nuclear power at the Atomic Energy Commission.

8.1.2 Authority and Responsibilities

8.1.2.1 Scope of Authority

The NRC's mission is to ensure that the civilian uses of nuclear energy and materials in the

United States are conducted with proper regard for public health and safety, national security,

environmental concerns, and, in the case of the initial licensing of nuclear power plants, the

antitrust laws. The Atomic Energy Act provides the charter for these regulatory responsibilities

through which Congress created a national policy of developing the peaceful uses of atomic

energy. Congress has amended the statute over the years to address developing technology

and changing perceptions of regulatory needs. For example, antitrust reviews were added in

1970, the same year that the National Environmental Policy Act of 1969, as amended, imposed

broad new responsibilities on Federal agencies. Other more specialized statutes prescribe the

45
NRC's duties with regard to high-level radioactive waste, low-level radioactive waste, mill

tailings, environmental reviews, nonproliferation, antiterrorism, and import/export.

The NRC's licensing authority extends to other government organizations (such as the

Tennessee Valley Authority (TVA), which operates nuclear power plants) and the military's use

of radiopharmaceuticals in its hospitals. The NRC's responsibilities include both safety arid

safeguards through which the agency ensures the security of commercial nuclear facilities and

materials against radiological sabotage and thefts.

In addition, the NRC is authorized to relinquish its authority, in certain cases, to States (i.e., of

the United States) who enter into agreements with the NRC (known as Agreement States).

Section 8.2 of this report provides specific information about the scope of the agency's authority

over DOE nuclear installations.

8.1.2.2 The NRC as an Independent Regulatory Agency

The Commission's status as an independent regulatory agency within ,the Executive Branch of

the Federal Government means that its regulatory decisions cannot ordinarily be directed by the

President. (By law, however, the U.S. Office of Management and Budget reviews the proposed

NRC budget.) Likewise, Congress cannot override the Commission's decisions, except by duly

enacted legislation.

The independence of the NRC's decisionmaking process implies a matching responsibility on

the part of the Commissioners and their personal staffs to keep the decisionmaking process free

from improper outside influence. 'This is especially important in the case of adjudications. When

the Commissioners take part in adjudications, they ordinarily act in the role of appellate judges

(reviewing the decisions of lower judges) and, in general, are bound by* the same kinds of

strictures that apply to judges in Federal courts.

8.1.3 Structure of the Regulatory Body

This section explains the structure of the NRC. It covers the Commission, component offices

and their responsibilities, and advisory committees and their functions. It also explains recent

changes in NRC organization.

8.1.3.1 The Commission

The NRC is headed by a five-member Commission. The President designates one member to

serve as Chairman and official spokesperson. The Commission as a whole formulates policies

and regulations governing nuclear reactor and materials safety, issues orders to licensees, and

adjudicates legal matters brought before it. The Executive Director for Operations carries out the

policies and decisions of the Commission and directs the activities of the program offices.

8.1.3.2 Component Offices of the Commission

Since the previous U.S. National Report was issued, the NRC has reorganized and established

two new offices-the Office of New Reactors and the Office of Federal and State Materials and

46
Environmental Management Programs. These offices, as well as other NRC offices, are briefly

described in the following paragraphs.

Office of the Executive Director for Operations. The Executive Director for Operations is the

chief operational and administrative officer of the Commission and is authorized and directed to

discharge such licensing, regulatory, and administrative functions and to take such actions as

necessary for day-to-day operations of the agency. The Executive Director supervises and

coordinates policy development and operational activities of NRC program and regional offices

and implements Commission policy directives pertaining to these offices.

Office of the Chief Financial Officer. The Office of the Chief Financial Officer is responsible for

the NRC's planning and budgeting and performance management process and for all NRC

financial management activities.

Office of the General Counsel. The Office of the General Counsel directs matters of law and

legal policy, providing opinions, advice, and assistance to the agency on all of its activities.

Office of the Inspector General. The Inspector General provides leadership and policy direction

in conducting audits and investigations to promote economy, efficiency, and effectiveness within

the NRC and to prevent and detect fraud, waste, abuse, and mismanagement in agency

programs and operations.

Office of International Programs. This office provides assistance and recommendations to the

Chairman, the Commission, and the NRC staff on international issues. It coordinates the NRC's

international activities; plans and implements policies in the international arena; and establishes

and maintains working relationships with individual countries and international nuclear

organizations as well as other involved U.S. Government agencies.

Office of Public Affairs. The Office of Public Affairs directs the agency's public affairs program,

advising agency officials and. developing key strategies that help increase public confidence in

NRC policies and activities.

Office of Congressional Affairs. The Office of Congressional Affairs is the primary point of

contact for all communications between the NRC and Congress. This office monitors legislative

proposals, bills, and hearings; informs NRC of the views of Congress on NRC policies, plans,

and activities; provides timely responses to congressional requests for information; and provides

the information necessary to keep appropriate Members of Congress and congressional staff

fully and currently informed of NRC actions.

The Office of Commission Appellate Adiudication. This office provides the Commission with an

analysis of any adjudicatory matter requiring a Commission decision and drafts necessary

decisions pursuant to the Commission's guidance after presentation of options.

Office of the Secretary of the Commission. This office provides executive management services

to support the Commission and to carry out Commission decisions. It assists with the planning,

scheduling, and conduct of Commission business; maintains historical paper files of official

Commission records; administers the NRC Historical Program; and maintains the Commission's

official adjudicatory and rulemaking dockets.

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8.1.3.3 Offices of the Executive Director for Operations

The offices reporting to the Executive Director for Operations ensure that the commercial use of

nuclear materials in the United States is safely conducted.

Office of Nuclear Reactor Regulation. The Office of Nuclear Reactor Regulation (NRR) is

responsible for accomplishing key components of the NRC's nuclear reactor safety mission. To

do so, NRR conducts a broad range of regulatory activities in the four primary program areas of

rulemaking, licensing, oversight, and incident response for commercial nuclear power reactors

and test and research reactors to protect the public health, safety, and the environment.

Office of New Reactors. This office is responsible for accomplishing key components of the

NRC's nuclear reactor safety mission for new reactor facilities licensed in accordance with

10 CFR Part 52. As such, NRO is responsible for regulatory activities in the primary program

areas of siting, licensing and oversight for new commercial nuclear power reactors.

Office of Nuclear Material Safety and Safeguards. The Office of Nuclear Material Safety and

Safeguards is responsible for regulating activities that provide for the safe and secure production

of nuclear fuel used in commercial nuclear reactors; the safe storage, transportation and

disposal of high-level radioactive waste and spent nuclear fuel; and the transportation of

radioactive materials regulated under the Atomic Energy Act.

Office of Nuclear Security and Incident Response. This office develops overall agency policy

and provides management direction for evaluating and assessing technical issues involving

security and emergency preparedness at nuclear facilities.

Office of Nuclear Regulatory Research. This office plans, recommends, and conducts research

programs to identify, lead, and sponsor reviews that support the resolution of ongoing and future

safety issues.

Regional Offices. The four regional offices conduct inspection, enforcement, and emergency

response programs for licensees within their borders.

Office of Enforcement. This office oversees, manages, and directs the development and

implementation of policies and programs for enforcing NRC requirements. It oversees the

agency's allegations management programs and the allegations review process. The office is

responsible for external safety culture policy matters, the agency's Alternative Dispute

Resolution program, the agency's internal Differing Professional Opinions Program, and its new

internal nonconcurrence process.

Office of Investigations. This office develops policy, procedures, and quality control standards

for investigations of licensees and applicants, as well as their contractors or vendors, including

the investigation of all allegations of wrongdoing by non-NRC employees and contractors.

Office of Federal and State Materials and Environmental Manaqement Programs. This office is

responsible for effective communications and working relationships between NRC and other

governmental entities and administers the Agreement State Program (through which States

have signed formal agreements with the NRC, to assume regulatory responsibility over certain

byproduct, source, and small quantities of special nuclear material). It also develops and

48
implements rules and guidance for the safe and secure use of source, byproduct and special

nuclear material in industrial, medical, academic, and commercial activities, and at

decommissioning, uranium recovery, and low-level waste sites.

This office was formerly the Office of State and Tribal Programs, but was reorganized to merge

with part of the Office of Nuclear Material Safety and Safeguards. This reorganization enhanced

the integration of the National Materials Program, a broad collective framework within which both

NRC and the Agreement States carry out their respective radiation safety regulatory programs.

The reorganization recognizes the increasing number of Agreement States, the value of their

experience in administering the National Materials Program, and the importance of coordination

between the NRC, the States, and other stakeholders.

Office of Information Services. This office plans, directs, and oversees the delivery of

centralized information technology infrastructure, applications, and information management

services, in addition to the development and implementation of information technology and

management plans, architecture, and policies to support the mission, goals, and priorities of the

agency.

Support Offices. Supporting the Executive Director of Operations are the Offices of

Administration, Human Resources, and Small Business and Civil Rights.

8.1.3.4 Advisory Committees

The three principal advisory committees for NRC programs are the Advisory Committee on

Reactor Safeguards, the Advisory Committee on Nuclear Waste, and the Advisory Committee

on Medical Uses of Isotopes. In addition, the NRC has established an ad hoc Licensing Support

Network Advisory Panel. The Advisory Committee on Reactor Safeguards is the only committee

relevant to this report. This committee reviews and reports on safety studies and reactor facility

license and license renewal applications; advises the Commission on the hazards of proposed

and existing reactor facilities and the adequacy of proposed reactor safety standards;; initiates

reviews of specific generic matters or nuclear facility safety-related items; and reviews the

NRC's research activities.

8.1.4 International Responsibilities and Activities

The NRC conducts international activities related to statutory mandates, international treaties,

conventions, codes, international organizations, bilateral relations, and research.

U.S. law or international treaties and conventions mandate several NRC international activities;

other activities are discretionary. In particular, the NRC is statutorily mandated to serve as the

U.S. licensing authority for exports and imports of nuclear materials and equipment.

The NRC supports U.S. foreign policies in the safe and secure use of nuclear materials and in

guarding against the spread of nuclear weapons. The agency actively participates in developing

and implementing a variety of legally binding treaties and conventions which create an

international framework for the peaceful uses of nuclear energy. The NRC provides technical

and legal advice and assistance to international organizations and foreign countries as they work

to develop effective regulatory organizations and rigorous safety standards. Some activities are

carried out within the programs of the IAEA, the Nuclear Energy Agency of the Organization for

49
Economic Cooperation and Development (NEA), or other international organizations. Others are

conducted directly with counterpart agencies in other countries under cooperation agreements.

Export-Import. The NRC's key international responsibility is licensing the export and import of

nuclear materials and equipment, such as low-enriched uranium fuel for nuclear power plants,

high-enriched uranium for research and test reactors, nuclear reactors themselves, certain

nuclear reactor components (such as pumps and valves), and radioisotopes used in industrial,

medical, agricultural and scientific fields. The NRC ensures that such exports and imports are

consistent with the goals of the safe and peaceful use of these materials and equipment, limiting

the proliferation of nuclear weapons, and promoting the Nation's common defense and security.

The Atomic Energy Act, the Nuclear Non-Proliferation Act of 1978, and 10 CFR Part 110,

"Export and Import of Nuclear Equipment and Material," detail the standards and procedures for

issuing export and import licenses.

International Treaties. Treaties that legally bind the NRC and the U.S. Government's peaceful

uses of nuclear energy include the 1978 Nuclear Non-Proliferation Treaty, the 1980 Convention

on Physical Protection of Nuclear Material, the 1994 Convention on Nuclear Safety, the 1986

Convention on Early Notification of a Nuclear Accident, the 1986 Convention on Assistance in

Case of a Nuclear Accident or Radiological Emergency, and the 1997 Joint Convention on the

Safety of Spent Fuel Management-and on the Safety of Radioactive Waste Management. NRC

staff regularly participate in international meetings related to these conventions and have held a

variety of Convention leadership positions. In its bilateral work with regulatory counterparts

worldwide, the NRC works to exchange experience and good practices in order to further the

goals of these international instruments.

International Orqanizations and Associations. The NRC actively participates in the full scope of

programs of the two major international nuclear organizations, the IAEA and the NEA. For

example, since 1996, the United States has or is planning to participate in more than 30

Operational Safety Assessment Review Team (OSART) missions. Some experts on these

teams come from the NRC, while others come from industry. The NRC coordinates closely with

INPO in this process. Since 1999, the NRC has participated in approximately 15 IRRT or IRRS

missions and intends to participate in at least four additional missions before the end of 2007.

The NRC is currently coordinating with the IAEA and industry in planning an OSART mission to

Arkansas Nuclear One Nuclear Power Plant in 2008 and intends to continue to plan for an

OSART mission in the United States every 3 years. The NRC also participates in the

Commission on Safety Standards and safety standards committees. Other IAEA activities

include the extra-budgetary program for long-term operation.

The NRC holds leadership roles in the four IAEA Safety Standards Committees and the

Commission on Safety Standards. These activities, together with regular NRC staff participation

in IAEA meetings to draft and revise safety and security guidance and coordination with other

U.S. Government agencies, enable the NRC to use its broad regulatory experience to contribute

to the safe and secure use of nuclear and radioactive materials in IAEA Member States. The

NRC also participates in the NEA Steering Committee, and holds leadership positions in the

Committee on the Safety of Nuclear Installations, the Committee on Nuclear Regulatory

Activities, the Committee on Radiation Protection and Public Health, and the Radioactive Waste

Management Committee. The NRC is also represented-on many of the NEA committeechartered working groups. These activities provide diverse fora for nuclear regulators and

research organizations to share information and work together to leverage resources for mutual

50
benefit.

The NRC continues to participate in the Multinational Design Evaluation Program (MDEP) with

the goal of leveraging the experience of international counterparts in the review of new reactor

designs. Currently, the NRC is participating in two parallel MDEP activities. In Stage 1 of the

MDEP, the NRC is cooperating with the-regulatory authorities in Finland and France on the

design reviews of the AREVA EPR. This reactor is currently being constructed in Finland, has

been approved for construction in France, and an application for design certification will be

submitted to the NRC in 2008. As part of Stage 2 of the MDEP, the NRC is participating with

nine other countries in a program to assess ways to enhance cooperation through the better

understanding and possible convergence of the country specific regulations, codes, and

standards associated with the design reviews of new reactors.

In addition to staff participation in more than 100 IAEA andNuclear Energy Agency meetings

each year, the NRC Chairman routinely participates in the IAEA General Conference and

biannual meetings of the International Nuclear Regulators Association. Members of the

Commission also travel to international conferences around the world to deliver keynote

remarks, participate in panel discussions, and otherwise share insight on a variety of topics to

diverse international audiences.

Bilateral Relations. The NRC works closely with nuclear safety agencies in more than 35

countries. The NRC and its foreign counterparts routinely exchange operational safety data and

other regulatory information. Subject to outside funding, the NRC provides safety, security,

emergency preparedness and safeguards advice, training, and other assistance to countries that

seek U.S. help to improve their regulatory programs.

The NRC's information exchange arrangements serve as communication channels with foreign

regulatory authorities, establishing a framework for the agency to gain access to non-U.S. safety

information which can (1) alert the U.S. Government and industry to potential safety problems,

(2) help identify possible accident precursors, and (3) provide accident and incident analyses,

including lessons learned, which could be directly applicable to the safety of U.S. nuclear power

plants and other facilities. The arrangements also serve as vehicles for the assistance the NRC

provides to countries to establish and improve their regulatory capabilities and infrastructure.

Thus, the arrangements facilitate NRC's strategic goal to support U.S. interests in the safe and

secure use of nuclear materials and in nuclear nonproliferation. There are currently 38 active

bilateral arrangements between the NRC and its foreign regulatory counterparts. These

arrangements allow the staff to conduct regular bilateral exchanges on a variety of levels. NRC

also has bilateral interactions with countries with which there is no active arrangement, though

the absence of an arrangement limits the type of information that can be exchanged. The NRC

Chairman typically meets with at least 20 foreign counterparts at the IAEA's annual General

Conference. In addition, members of the Commission travel abroad to hold bilateral meetings

with their regulatory counterparts, tour nuclear power plants and other facilities, and exchange

information and good practices. Often, these visits result in increased communication between

the NRC and-its counterparts, providing opportunities for enhanced information exchange based

on first-hand knowledge of various programs.

International Assistance Programs. In the early 1990s, the NRC began offering assistance to

nuclear regulatory programs in several former Soviet states. Efforts were initially focused on

those countries in which Soviet-designed reactors were operated. Following the September 11,

51
2001 terrorist attacks, assistance efforts were expanded to specifically include assisting

countries in their efforts to improve regulatory oversight of radioactive sources. In addition, the

NRC is assisting the Government of Iraq in its efforts to develop a sound regulatory structure,

including the provision of assistance in developing the law and regulations that will be the legal

frameworkfor the project of decommissioning former facilities that used radioactive materials in

Iraq.

Research Programs. The NRC conducts confirmatory regulatory research through more than 90

multilateral agreements in partnership with nuclear safety agencies and institutes in more than

20 countries. This research supports regulatory decisions on emerging technologies, aging

equipment and facilities, and various other safety issues. The NRC and other nuclear regulatory

and safety organizations carry out cooperative research projects to achieve mutual research

needs with greater efficiency. This research is currently conducted under the following

programs: Cooperative Severe Accident Research Program; Code Applications and

Maintenance Program; and Steam Generator Tube Integrity Program.

8.1.5 Financial and Human Resources

8.1.5. 1 Financial Resources

As of September 30, 2006, and 2005, the financial condition of the NRC was sound with respect

to having sufficient funds to meet program needs and adequate control of these funds in place

to ensure obligations did not exceed budget authority. The sum of all funds available to obligate

for FY 2006 was $809.0 million, which is a $86.1 million increase over the FY 2005 amount of

$722.9 million.

The NRC FY 2008 budget will be financed with $765.1 million from user fees, $114.3 million

from the General Fund, and $37.3 million from the Nuclear Waste Fund.

8.1.5.2 Human Resources

The NRC was ranked as the best place to work in the federal government among large federal

agencies in 2007. This ranking, along with new recruiting authority provided by Congress,

should assist in the agency's hiring efforts to maintain an innovative and effective workforce.

The NRC is recruiting about 400 employees each year for the next few years because of the

expected arrival of close to two dozen applications for new reactor licenses beginning in the fall

of 2007. The NRC is working to aggressively meet the challenge of recruiting, hiring, training

and integrating new employees with the necessary infrastructure to support its activities, as well

as retaining staff, as discussed below.

Recruitment and Hiring Process. To meet the demands of hiring new employees and to address

the need for more experienced individuals, the NRC identified the need to expand its recruitment

activities and streamline the hiring process. As a long standing practice, NRC actively recruits

for its Nuclear Safety Professional Development Program at targeted universities with a history

of graduating technically strong, diverse candidates. It has expanded its recruitment activities at

professional society conferences and career fairs, and is expanding advertising in trade journals

and on Web sites to attract professionals in specialized technical disciplines and in local

newspapers around the country where technical engineers and scientists may be interested in

re-locating due to job cutbacks in their areas.

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The Agency has streamlined recruitment, relocation, and retention incentives to allow offices to

extend job and incentive offers to outside applicants. For example, NRC has revised the generic

open vacancy announcement for mid-career engineers and scientists to provide additional

flexibilities to offer relocation and recruitment incentives. The NRC is also, creating the policy of

offering referral awards. Other innovations, such as student loan repayments, waivers of dual

compensation limitations, partnerships with colleges and universities, and the Cooperative

Education Program, have an equally positive impact on the Agency's efforts to recruit and retain

staff with critical skills.

Training and Knowledge Management (KM)/transfer. About 40% of staff have been with the

NRC for less than 4 years. The rapid integration and training of such a large number of new

employees into the Agency is a significant challenge but is essential for the NRC's and the

employees' future success and productivity. To address this challenge, the NRC is expanding

the use of existing training tools, including mentoring, on-the-job-training, formal classroom and

on-line training, and self-study activities. Senior staff are being asked to help train newer staff in

addition to their regular workload and to help new employees assimilate into the NRC culture. A

major challenge is the multi-generational population now working together, each with different

ways of approaching work.

NRR developed an integrated approach to training to provide new employees with consistent

information from branch to branch and division to division. NRR developed a qualification

program that consists of three parts: general requirements, position specific requirements, and

an oral confirmation board. The training is focused on developing engineers and scientists into

regulators. To ensure NRR's administrative staff continues to excel, an administrative support

training program was developed.

To help new employees adjust to the NRC, an "NRR New Employee Orientation and Training

Guide" has been developed and is being implemented. The guide includes training courses,

reading assignments and self-study activities. Additionally, new employees will be assigned a

"docent" that will be a peer, typically from their branch, to assist in adjusting. Additionally,

position-specific training is expected to accompany the generic training in the "NRR New

Employee Orientation and Training Guide." The NRC is continuing to develop its qualification

plans and other position-specific training, for example, for project engineers/managers. It is also

identifying course needs at its Technical Training Center and Professional Development Center.

For succession planning and knowledge management in critical skills and knowledge areas,

supervisors and managers have been provided a new tool in the NRC's Strategic Workforce

Planning system to help identify skill gaps. Using the skill categories and the needs assessment

existing in the system, supervisors can view a staff and critical skill matrix. This matrix displays

the branch employee's level of expertise in each of the most critical skills identified by the branch

chief. By identifying potential skill gaps, supervisors can make more informed decisions when

assigning work, and can identify skill areas for individual employee development.

Workforce Planning and Deployment With a renewed emphasis on hiring to meet the expected

increase in new reactor work, NRR realigned to emphasize the area of new reactors, and the

Office of Nuclear Material Safety and Safeguards realigned to enhance cooperation with States

and implement a holistic approach to fuel issues including transportation, storage, and disposal.

The changes in these two offices were made easier by the NRC's strategic workforce planning

tool which allowed for a smoother planning process to improve workforce deployment, maintain

53
technical capacity, and make informed decisions on human capital strategies for recruitment,

development, and retention.

Leadership and KM. The NRC continues to offer and expand its leadership competency

development programs, such as executive leadership seminars, the Senior Executive Service

Candidate Development Program, leadership training for new supervisors and team leaders, and

the Leadership Potential Program.

KM is a part of the strategic management of human capital, along with strategic workforce

planning, recruitment, and training and development. As part of this effort, the NRC is in the

process of coordinating its efforts to implement KM strategies, including developing a KM Web

site.

In addition, the NRC is developing an Agencywide KM plan that will serve as a framework to

integrate new and existing approaches that generate, capture, and transfer knowledge and

information relevant to the NRC's mission. The following are some of the near-term and longterm strategies for this plan:

• capture relevant critical knowledge of departing personnel

" recapture departed knowledge where possible

" communicate leadership's expectation for a knowledge-sharing culture

* formalize KM values and principles

" incorporate KM within process work flows

Following are some of the current Knowledge Management/Knowledge Transfer Activities.

Branch Chief/Team Leader Seminars. As the role and expectations of branch chiefs

evolved from senior technical experts to managers, it is essential that they have the

information they need to succeed in their positions. As a community of practice, the

branch chiefs/team leaders meet monthly, and are given presentations by agency

experts in topics such as performance management, budget, and communications.,

Branch/Team Meetings. To ensure staff in each- branch or team are kept up-to-date in

areas under their purview, branch chiefs and team leaders hold regularly scheduled staff

meetings. During some of these meetings, senior staff members are asked to give

presentations to staff regarding an area in which they are considered experts, or passing

their knowledge of past events on to newer staff. Some branch chiefs also have their

more junior staff give presentations. This is also a way for junior and senior staff to

interact, as the junior staff member may be required to interview more senior staff

members to glean information for their presentations.

Video Interviews. NRC conducted a pilot to capture.knowledge from retiring senior staff

using video interviews. One video captured knowledge regarding steam generators;

another was entitled "Nuclear Knowledge for the Next Generation." The interviews

included questions about licensing issues, recruiting and mentoring new hires,

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leadership, operations center experience, and reactor licensing performance metrics.

Web Sites NRR has a very active web site devoted to KM. Entitled "Sharing Expert

Experience and ,Knowledge" (SEEK), this site contains information such as the Inspector

Best Practices Booklet and Inspector Newletters, supervisor and team leader seminars,

new employee orientation and training guide, key reference materials for reviews,

qualification plans, strategic workforce planning, KM, and other communities of practice.

Communities of Practice. NRR is piloting an electronic community of practice

Demonstrations have been given to a technical, regulatory and policy consistency

working group for consideration. Demonstrations to other groups who may benefit by its

use continue to be given.

Retaininq Staff. Ability to retain recent recruits as opportunities arise in the industry is a

challenge, because people are no longer committed to working in one place for their entire

career. However, challenging work alone does not ensure retention. The NRC believes that the

benefits of the workplace will more than likely help in retaining staff. To this end, NRC

continues to try to create an environment where people feel valued and challenged. For

example, NRC has made work hours flexible and the ability to work out in the on-site fitness

center to help staff create a balance in their life. It has created an organization that the staff

feels a part of; one that lives up to its values and has an open environment where

communication flows in all directions, and challenging the status quo is encouraged.

8.1.6 Position of the NRC in the Governmental Structure

This.section explains the relationship of the NRC to the Executive Branch, the States, and

Congress.

8.1.6.1 Executive Branch

The components of the Executive Branch with which the NRC has the most frequent contact and

interaction are the White House, Office of Management and Budget, Department of State, DOE,

EPA, FEMA, Department of Labor, Department of Transportation, and Department of Justice.

Section 8.2 of this report discusses the NRC's relationship to DOE. The agency's relationships

with other components of the Federal Government are described below.

The White House. As noted above, as an independent regulatory agency, the White House

cannot directly set NRC policy. It can, however, influence NRC policy by (1) appointing

Commissioners and Chairmen in whose outlook and judgment it has confidence and (2) making

its views known on nonadjudicatory matters. In certain areas, such as national security policy,

the Commission has declared its intent to give great Weight to the views of the Executive

Branch. In informal policy-matters, such as rulemaking, White House and Executive Branch

officials may properly try to influence NRC decisions, either publicly or privately. Ultimately,

however, the NRC must make the decision and accept responsibility for it.

Office of Management and Budget. The NRC submits its annual budget requests, including

proposed personnel ceilings, to the Office of Management and Budget for approval.

U.S. Department of State. By law, the NRC must license the export and import of nuclear

55
equipment and material. For significant applications, the Commission requests the U.S.

Department of State to provide Executive Branch views on whether the license should be issued.

The NRC also works with the State Department negotiating international agreements in the

nuclear field and interacting with the IAEA and other international organizations of the United

Nations, as well as the Nuclear Energy Agency of the Organization for Economic Cooperation

and Development. In general, the purposes of these interactions are to develop policy on

nuclear issues that are under NRC purview and to plan and coordinate, programs of nuclear

safety and safeguards assistance to other countries.

EPA. The responsibilities of NRC and EPA intersect or overlap in areas in which the EPA issues

generally applicable environmental standards for activities that are also subject to NRC

licensing. Examples include standards for high-level waste repositories, decommissioning

standards, and standards for public and worker protection. The EPA has the ultimate authority

to establish generally applicable environmental standards to protect the environment from

radioactive material.

FEMA FEMA assists the NRC's licensing process by preparing reviews and evaluations, as well

as presenting witnesses to testify at licensing hearings. FEMA also participates with the NRC in

observing and evaluating emergency exercises at nuclear plants. FEMA findings are not binding

on the NRC, but they are presumed to be valid unless controverted by more persuasive

evidence. FEMA is now part of DHS.

Department of Transportation. The NRC and the Department of Transportation share

responsibility for the control of radioactive material transport. Department of Transportation

regulations cover all aspects of transportation, including packaging, shipper and carrier

responsibilities, documentation, and all levels of radioactive material.

U.S. Department of Labor. The NRC monitors discrimination actions related to NRC-licensed

activities filed with the Department of Labor under Section 211 of the Energy Reorganization Act

and develops enforcement actions when there are properly supported findings of discrimination,

either from the NRC's Office of Investigations or from U.S. Department of Labor adjudications.

U.S. Department of Justice. Any NRC litigation almost always requires coordination with the

U.S. Department of Justice. Under the Administrative Orders Review Act (commonly called the

Hobbs Act), the United States is a party to petitions for review challenging NRC licensing

decisions or regulations.

The Office of Investigations, which investigates allegations of wrongdoing by NRC applicants

and licensees, as well as by their contractors, normally works with the General Litigation Section

of the Criminal Division at the Department's Headquarters andwith U.S. Attorneys in the field.

The Office of the Inspector General reports to the Department of Justice whenever it has

reasonable grounds to believe that an NRC employee or contractor has violated Federal law.

The Inspector General refers cases for review for possible criminal prosecution to the U.S.

Attorney's Office for the area in which the potential violation occurred. When the Department of

Justice desires support from the Office of the Inspector General for investigations or grand jury

work, it makes the request directly to the Inspector General.

56
8.1.6.2 The States (i.e., of the United States)

At the NRC, the Office of Federal and State Materials and Environmental Management

Programs is responsible for establishing and maintaining effective communications and working

relationships between the NRC and the States and serves as the primary contact for policy

matters, keeping the States informed about NRC activities and keeping the NRC appraised of

State activities and views that may affect NRC policies, plans, and activities. Other NRC offices

provide major support to implement State relations program policy and guidance, for example,

through regional state liaisons and State agreements officers.

As explained above, the Atomic Energy Act confers on NRC preemptive authority over health

and safety regulation of nuclear energy and Atomic Energy Act materials. As a result, the

general rule is that nuclear power plant safety, like airline safety, is the exclusive province of the

Federal Government and cannot be regulated by the States. A State law that attempted to set

nuclear safety standards would thus be voided by the courts. However, the courts will not

overturn a State law that regulates nuclear energy for purposes other than health and safety,

such as economics, unless it conflicts with an NRC requirement. Similarly, the courts will not

ordinarily question a State's declared purpose in enacting legislation.

However, the Atomic Energy Act did not entirely exclude States from the regulation of nuclear

matters. Section 274 of the act created the Agreement State Program, under which the NRC

may relinquish its authority over most nuclear materials to those States willing to assume that

authority. The NRC may not relinquish authority over such facilities as reactors, fuel

reprocessing and enrichment plants, imports and exports, critical mass quantities of special

nuclear material, high-level waste disposal, or certain other excepted areas.

Many States have signed formal agreements with the NRC and have assumed regulatory

responsibility over certain byproduct, source, and small quantities of special nuclear material.

Agreement States receive no Federal funding to support their regulatory programs. The NRC

conducts performance-based reviews of Agreement State programs to ensure that they remain

adequate to protect the public health and safety and are compatible with the NRC materials

program.

Some States have shown a desire to participate in matters relating to nuclear power plants.

In response, the NRC issued a policy statement in February 1989 declaring its intent to

cooperate with States in the area of nuclear power plant safety by keeping States informed of

matters of interest to them and considering proposals for State officials to participate in NRC

inspection activities, pursuant to a memorandum of understanding between the State and NRC.

The policy statement makes clear that States must channel their contacts with the NRC through

a single State Liaison Officer, appointed by the Governor. States are authorized only to observe

and assist in NRC inspections of reactors and they cannot conduct their own independent health

and safety inspections.

Through the Intergovernmental Liaison Program, the NRC works in cooperation with Federal,

State, and local governments; interstate organizations; and Native American Tribal Governments

to maintain effective relations and communications with these organizations and to promote

greater awareness and mutual understanding of the policies, activities, and concerns of all

parties involved as they relate to radiological safety at NRC-licensed facilities.

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8.1.6.3 Congress

The following paragraphs discuss the NRC oversight committees in the Senate and House and

subcommittees with jurisdiction over aspects of NRC's activities.

Senate Oversight. In the U.S. Senate, the Committee on the Environment and Public Works has

jurisdiction over domestic nuclear regulatory activities. With the committee, the Subcommittee

on Clean Air and Nuclear Safety has responsibility for regulation and oversight of the NRC. The

Energy and Natural Resources Committee and the Environment and Public Works Committee

share jurisdiction over nuclear waste issues.

House Oversight. In the U.S. House of Representatives, the Committee on Energy and

Commerce has jurisdiction over domestic nuclear regulatory activities. Within the committee,

the Subcommittee on Energy and Air Quality has responsibility for regulation and oversight of

the NRC.

Other Relevant Committees. In addition to the committees and subcommittees mentioned

above, the House and Senate Appropriations Subcommittees on Energy and Water

Development play a key role in approving the Commission's annual budget. Also, a number of

committees frequently interface with the NRC concerning international affairs, research, security,

and general governmental operations.

8.1.7 Report of the IRRS Self-Assessment Team

At the third review meeting of the Convention, the U.S. committed to perform an IRRT (now

called the IRRS) self-assessment.

In the fall of 2006, an inter-office IRRS self-assessment team of about 14 people was

constituted. The team consisted of representatives from the Offices of: Nuclear Reactor

Regulation, Nuclear Regulatory Research, Federal and State Materials and Environmental

Management Programs, Nuclear Materials Safety and Safeguards, Nuclear Security and

Incident Response, General Counsel, Chief Financial Officer, Human Resources and

International Programs.

The IRRS self-assessment questionnaire was developed in collaboration with the Canadian

Nuclear Safety Commission and the IAEA and based on the IRRT questionnaire (IAEA

TECDOC-703) and IAEA draft document entitled "Regulatory Performance Self-Assessment

Tool." Supplemental questions were added to help assess the regulatory readiness for potential

new reactor licensing. The IRRS self-assessment questionnaire consists of over 275 questions.

The team reviewed the results of the IRRS self-assessment responses and developed 12 highlevel recommendations. These recommendations are being reviewed by a senior management

review board, which will propose appropriate staff treatment in response to them.

8.2 Separation of Functions of the Regulatory Body from Those of Bodies

Promoting Nuclear Energy

Although both the NRC and DOE have responsibilities for managing nuclear facilities and

materials, they maintain separate, independent functions. The partitioning of the U.S. Atomic

58
Energy Commission in the mid-1 970s provided distinct entities for the U.S. Government's

regulatory and promotional responsibilities in nuclear applications.

Specifically, the Energy Reorganization Act redistributed the functions performed by the U.S.

Atomic Energy Commission to two new agencies. This act created the NRC to regulate the

commercial nuclear power sector and ERDA to promote energy and nuclear power development

and to develop defense applications. The NRC was established as an independent authority to

regulate the possession and use of nuclear materials as well as the siting, construction, and

operation of nuclear facilities. ERDA was established to ensure the.development of all energy

sources, increase efficiency and reliability of energy resource use, and carry out the other

functions, including but not limited to the U.S. Atomic Energy Commission military and production

activities and general basic research activities.

The NRC performed its regulatory mission by issuing regulations, licensing commercial nuclear

reactor construction and operation, licensing the possession of and/or use of nuclear materials

and wastes, safeguarding nuclear materials and facilities from theft and radiological sabotage,

inspecting nuclear facilities, and enforcing regulations. The NRC regulates the commercial

nuclear fuel cycle materials and facilities. Regarding the regulatory control of commercial spent

nuclear fuel and radioactive waste, the NRC is responsible for licensing commercial nuclear

waste management facilities, independent spent fuel management facilities, and the DOEproposed Yucca Mountain site for the disposal of high-level waste and spent fuel.

DOE addresses the U.S. Government's need to unify energy organization and planning. The

DOE Organization Act brought a number of the Federal Government's agencies and programs,

including ERDA, into a single agency with responsibilities for nuclear energy technology and

nuclear weapons programs. Over the past decade, DOE has added new nuclear-related

activities directed to environmental cleanup of contaminated sites and surplus facilities and

nonproliferation. DOE retains authority under the Atomic Energy Act for regulating its nuclear

activities. DOE also retains responsibility for regulating the disposal of its own low-level

radioactive waste.

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;1
ARTICLE 9. RESPONSIBILITY OF THE LICENSE HOLDER

Each Contracting Party shall ensure that prime responsibility for the safety of a nuclear

installation rests with the holder of the relevant license and shall take the appropriate

steps to ensure that each such license holder meets its responsibility.

The NRC, through the Atomic Energy Act, ensures that the prime responsibility for the safety of

a nuclear installation rests with the licensee. Steps the NRC takes to ensure that each licensee

meets its primary responsibility include the licensing process, discussed in Articles 18 and 19,

the Reactor Oversight Process, discussed in Article 6, and the Enforcement Program, discussed

below. This update revises the debt collection dollar amount, discusses the alternative dispute

resolution program and current experience.

9.1 Introduction

The NRC's regulatory programs continue to be based on the premise that the safety of

commercial nuclear power reactor operations is the responsibility of NRC licensees. The

responsibility of the NRC is regulatory oversight of licensee activities to ensure that safety is

maintained. The NRC reviews the safety of a reactor design and the capability of an applicant to

design, construct, and operate a facility.. If an applicant satisfies the requirements of the Code of

Federal Regulations, the NRC then issues a license to operate the facility. Such licenses specify

the terms and conditions of operation to which a licensee must conform. Failure to conform

subjects the licensee to enforcement action, which can include modifying, suspending, or

revoking the license. The NRC can also order particular corrective actions or issue civil

penalties. The following paragraphs discuss these enforcement mechanisms in greater detail.

9.2 The Licensee's Prime Responsibility for Safety

As discussed in Article 7 of this report, Chapter 10, Section 103 of the Atomic Energy Act grants

the NRC authority to issue licenses for nuclear reactor facilities. Moreover, Section 103 states

that these licenses are subject to such conditions as the NRC may by rule or regulation establish

to effectuate the purposes and provisions of the Atomic Energy Act. Consistent with the act,

before issuing a license, the Commission determines that the applicant is (1) equipped and

agrees to observe such safety standards to protect health and to minimize danger to life or

property as the Commission may by rule establish and (2) agrees to make available to the

Commission such technical information and data about activities under such license as the

Commission may determine necessary to promote the common defense and security and to

protect the health and safety of the public.

Embedded in each license is the explicit responsibility that the license holder comply with the

terms and conditions of the license and the applicable Commission rules and regulations. The

licensee is ultimately responsible for the safety of its activities and the safeguarding of nuclear

facilities and materials used in operation.

When the Commission or licensee determines that the licensee is not complying with the

Commission's rules or regulations, action is taken to ensure that the facility is returned to a

condition compliant with its license.

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9.3 NRC Enforcement Program

As discussed in Article 7, the NRC has enforcement powers. As discussed in Section 6.2.2.5,

the enforcement process complements the Reactor Oversight Process. The NRC uses

enforcement as a deterrent to emphasize the importance of compliance with regulatory

requirements and to encourage prompt identification and prompt, comprehensive correction of

violations.

The NRC identifies violations through inspections and investigations. All violations are subject to

civil enforcement action and may be subject to criminal prosecution. Unlike the burden of proof

standard for criminal actions (beyond a reasonable doubt), the NRC uses the Administrative

Procedure Act standard (preponderance of evidence) in enforcement proceedings. After an

apparent violation is identified, it is assessed in accordance with the Commission's enforcement

policy, found in NUREG-1600, "General Statement of Policy and Procedures for NRC

Enforcement Actions," issued July 2000, which is available to NRC licensees and members of

the public. The NRC's Office of Enforcement maintains the current policy statement on NRC's

public Web site. Because it is a policy statement and not a regulation, the Commission may

deviate from it, as appropriate, given the circumstances of a particular case.

The NRC has three primary enforcement sanctions available. Specifically, those sanctions

include notices of violation, civil penalties, and orders.2 A notice of violation identifies a

requirement and how it was violated, formalizes a violation pursuant to 10 CFR 2.201, "Notice of

Violation," requires corrective action, and normally requires a written response. A civil penalty is

a monetary fine issued under authority of Section 234 of the Atomic Energy Act or Section 206

of the Energy Reorganization Act. Section 234 of the Atomic Energy Act provides for penalties

of up to $100,000 per violation per day; however, that amount is adjusted every 4 years by the

Federal Civil Penalties Inflation Adjustment Act of 1990, as amended by the Debt Collection

Improvement Act of 1996, and is currently $130,000. The Commission's order-issuing authority

under Section 161 of the Atomic Energy Act is broad and extends to any area of licensed activity

that affects public health and safety or the common defense and security. Orders modify,

suspend, or revoke licenses, or they may require specific actions by licensees or persons. The

NRC issues notices of violatio ns and civil penalties on the basis of violations. Orders may be

issued for violations or, in the absence of a violation, because of a public health or safety or

common defense and security issue.

After identifying a violation, the NRC assesses its significance by considering the following

factors:

0 actual safety consequences

0 potential safety consequences

0 potential for impacting the NRC's ability to perform its regulatory function

0 any willful aspects of the violation

Given those factors, the NRC takes one of the following actions based on the significance of the

violation:

The NRC also uses administrative actions, such as notices of deviation, notices of nonconformance,

confirmatory action letters, and demands for information to supplement its enforcement program.

62
assigns a severity level, ranging from Severity Level IV (more than minor concern) to

Severity Level I (the most significant)

associates the violation with findings assessed through the Reactor Oversight Process

significance determination. process (described in Article 6) and assigns a color code of

green, white, yellow, or red based on increasing risk significance

The Commission recognizes that there are violations of minor safety or environmental concern

.that are below Severity Level IV violations, as well as below violations associated with green

findings. These minor violations are not assigned a severity level category or a color

assessment.

The NRC may hold a predecisional enforcement conference or a regulatory conference with a

licensee before making an enforcement decision if (1) escalated enforcement action appears

warranted, (2) the NRC decides a conference is necessary, or (3) the licensee requests it.

The purpose of the conference is to obtain information. to assist the NRC in determining the

appropriate enforcement action, such as a common understanding of facts, root causes, and

missed opportunities associated with the apparent violations; corrective actions taken or

planned; and the significance of issues and the need for lasting, comprehensive corrective

actions.

At several junctions during the enforcement process involving cases of discrimination or willful

violation of NRC regulations, the agency offers its licensees or individuals the opportunity to

participate in the Alternative Dispute Resolution Program. Alternative dispute resolution is a

general term encompassing various techniques for resolving conflict outside of court using a

neutral third party. The NRC uses mediation, a technique in which a neutral mediator with no

decisionmaking authority helps parties clarify issues, explore settlement options, and evaluate

how best to advance their respective interests. Neutral mediators are selected from a roster of

experienced mediators provided by a neutral program administrator who is'under contract with

the NRC. The mediator's responsibility is to assist the parties in reaching an agreement.

However, the mediator has no authority to impose a resolution upon the parties. Mediation is a

confidential and voluntary process. If the parties to the process (the NRC and the licensee or

individual) agree to use alternative dispute resolution, they select a mutually agreeable neutral

mediator and share equally the cost of the mediator's services. In cases in which the NRC and

the licensee or the individual reach an agreement, the agency issues a confirmatory order

reflecting the terms of the agreement.

The agency normally assesses civil penalties for Severity Level I and II violations, as well as

knowing and conscious violations of the reporting requirements of Section 206 of the Energy

Reorganization Act. Civil penalties are considered for Severity Level III violations. Although not

normally used for violations associated with the Reactor Oversight Process, civil penalties

(and the use of severity levels) are considered for issues that are willful, that have the potential

to affect the regulatory process, or that have actual consequences.

Although each severity level may have several associated considerations, the outcome

of the assessment process for each violation or problem (absent the exercise of discretion)

results in one of three outcomes, which may involve no civil penalty, a base civil penalty, or twice

the base civil penalty.

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The NRC may issue orders to modify, suspend, or revoke a license; issue orders to cease and

desist from a given practice or activity; or take such other action as may be proper. The agency

may issue orders in lieu of, or in addition to, civil penalties. Additionally, the NRC may issue an

order to impose a civil penalty when a licensee refuses to pay a civil penalty or an order to an

unlicensed person (including vendors) when the agency has identified deliberate misconduct.

By statute, a licensee or individual may request a hearing upon receiving an order. Orders are

normally effective after a licensee or individual has had an opportunity to request a hearing

(i.e., 30 days). However, orders can be made immediately effective without prior opportunity for

a hearing when the agency determines it is the best interest of public health and safety to do so.

Subsequent to the hearing process, a licensee or individual may appeal the administrative

hearing decision to the Commission and, if desired, appeal the Commission's decision to a U.S.

court of appeals.

Providing interested stakeholders with enforcement information is very important to the NRC.

Conferences that are open to public observation are included in the listing of public meetings

on NRC's public Web site. The agency issues a press release for each proposed civil penalty or

order. All orders are published in the Federal Register. Significant enforcement actions

(including actions to individuals) are included in the enforcement document collection in the

Electronic Reading Room of NRC's public Web site.

From October 1, 2004, through September 30, 2005, the NRC issued a variety of significant

enforcement actions to operating power reactors. Specifically, these actions included 23

escalated notices of violation without civil penalties, ten civil penalties, one confirmatory order,

and one order imposing a civil monetary penalty.

During 2006, NRC issued a variety of significant enforcement actions to operating power

reactors. Specifically, these actions included 26 escalated notices of violation without civil

penalties, three civil penalties, and two orders.

To provide accurate and timely information to all interested stakeholders and enhance the

public's understanding of the enforcement program, the NRC publishes related information on

NRC's public Web site.

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ARTICLE 10. PRIORITY TO SAFETY

Each Contracting Party shall take the appropriate steps to ensure that all organizations

engaged in activities directly related to nuclear installations shall establish policies that

give due priority to nuclear safety.

NRC policies that give due priority to safety covered under this article are PRA policy statements

and policies that apply to licensee safety culture and safety culture at the NRC.

Other articles (e.g., Articles 6, 14, 18, and 19) also discuss activities undertaken to achieve

nuclear safety at nuclear installations.

This section was updated to discuss new regulations, developments in PRA, and safety culture.

10.1 Background

The United States has made substantial progress in developing and using the results of PRAs

for all operating reactor facilities, and the NRC has developed extensive guidance regarding the

role of PRA in U.S. regulatory programs. The NRC has extensively applied information gained

from PRA to complement other engineering analyses in improving issue-specific safety

regulation and in changing the current licensing bases for individual plants. The move toward

risk-informing the current regulations and processes continues to mark perhaps the most

significant changes at the NRC. An example since the third review meeting is 10 CFR 50.69,

"Risk-Informed Categorization and Treatment of Structures, Systems, and Components,"

promulgated on November 22, 2004. This new, voluntary rule modifies the scope of the special

treatment regulations in 10 CFR Part 50. It does so by creating an alternative regulatory

framework that enables licensees to use a risk-informed approach to categorize SSCs, and their

associated treatment, according to their safety significance. The NRC is also continuing a

program to develop additional changes to the specific technical requirements in the body of

10 CFR Part 50.

The NRC is considering an approach that, in addition to the ongoing effort to revise some

specific regulations to be risk-informed and performance-based, would establish a

comprehensive set of risk-informed and performance-based requirements applicable for all

nuclear power reactor technologies as an alternative to current requirements. This new rule (10

CFR Part 53) would take advantage of operating experience, lessons learned from the current

rulemaking activities, advances in the use of risk-informed technology, and would focus NRC

and industry resources on the most risk-significant aspects of plant operations to better ensure

public health and safety. The set of new alternative requirements would be intended primarily

for new power reactors although they would be available to existing reactor licensees. On May

4, 2006, the NRC issued an Advanced Notice of Proposed Rulemaking seeking stakeholder

input and currently is evaluating that input.

10.2 Probabilistic Risk Assessment Policy

Three policy statements form the b6asis for the NRC's current treatment of PRA and the related

regulatory safety goals and objectives-the "Policy Statement on Severe Reactor Accidents

Regarding Future Designs and Existing Plants," issued August 8, 1985; the "Safety Goals for the

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Operations of Nuclear Power Plants Policy Statement," issued August 21, 1986; and the "Policy

Statement on Use of PRA Methods in Nuclear Activities," issued August 16, 1995. Previous U.S

National Reports have detailed these policies.

10.3 Applications of Probabilistic Risk Assessment

The NRC applies PRA to resolve severe accident issues, evaluate new and existing

requirements and programs, implement risk-informed regulation, and improve data and methods

of risk analysis. The NRC also engages in cooperative activities with industry (such as pilot

programs for 10 CFR 50.69, 10 CFR 50.48(c), and Regulatory Guide 1.200) and in activities that

assess risk in determining plant-specific changes to the licensing basis.

The NRC maintains a risk-informed and performance-based plan (RPP), which sets forth the

agency's planned actions to risk-inform and performance-base its regulatory activities. In the

past, the Risk-Informed Regulation Implementation Plan focused largely on risk-informed

initiatives. In the current improved plan, the objectives have been expanded to more fully

achieve a risk-informed and performance-based regulatory structure. A website will be

developed for the RPP with links to documents that specifically describe activities and status.

The NRC and industry representatives have cooperated in a number of activities and pilot

programs to develop and apply risk-informed methodologies for specific regulatory applications.

The staff uses the lessons learned from these activities to enhance the effectiveness of

developed guidance. The activities described in this section are inservice testing, inservice

inspection, technical specification changes, and standards development.

10.3.1 Risk-Informed Inservice Testing

The agency has approved or is reviewing several applications of risk-informed inservice testing,

of generally limited scope. For example, in August 2001, the staff granted a risk-informed

exemption request from the licensee of the South Texas Project regarding special treatment

requirements of low-risk and nonrisk-significant safety-related nuclear components (including an

exemption from prescriptive inservice testing requirements). Having successfully implemented

this exemption, the staff developed a new rule, 10 CFR 50.69 (discussed in Section 10.1 of this

report), to allow risk insights to be applied to reduce the special treatment requirements in,

10 CFR Part 50 for SSCs that are categorized as being of low safety significance.

As another example, the NRC amended 10 CFR 50.55a, "Codes and Standards," by publishing

a final rule, entitled "Incorporation by Reference of ASME BPV and OM Code Cases." This

rulemaking incorporated by reference specific revisions of NRC Regulatory Guides 1.84,

"Design, Fabrication, and Materials Code Case Acceptability," issued March 2003; 1.147,

"Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," issued January

2004; and 1.192, "Operations and Maintenance Code Case Acceptability, ASME OM Code,"

issued June 2003, which list the ASME Code Cases that the NRC accepts as alternatives to

ASME Code requirements. Since Regulatory Guide 1.192 lists acceptable (and conditionally

acceptable) OM Code Cases, including risk-informed categorization and component-specific

code cases, licensees can now implement risk-informed inservice testing programs without

following Regulatory Guide 1.175 and without prior NRC approval.

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10.3.2 Risk-Informed Inservice Inspection

The NRC uses the guidance in Revision 1 to Regulatory Guide 1.178, "An Approach for Plant

Specific Risk-Informed Decision-making for Inservice Inspection of Piping," issued September

2003, and the corresponding Section 3.9.8 of NUREG-0800. The agency-approved industry

methodologies, developed by the Westinghouse Owners Group and the other by EPRI,

regarding alternatives to the ASME Xl Inservice Inspection Program continue to be used for

inservice inspections.

The NRC regularly participates in the ASME Code development process to resolve issues.

regarding risk-informed inservice inspection methodology.

10.3.3 Risk-Informed Technical Specification Changes

Since the mid-1980s, the NRC has reviewed and granted improvements to technical

specifications that are based, at least in part, on PRA insights. In its final policy statement on

technical specification improvements published on July 22, 1993, the Commission stated that it

expects licensees to use any plant-specific PRA or risk survey in preparing submittals related to

technical specifications. The Commission reiterated this point when it revised 10 CFR 50.36,

"Technical Specifications," in July 1995.

The NRC continues to use Regulatory Guide 1.177, "An Approach for Plant-Specific, RiskInformed Decisionmaking: Technical Specifications," issued August 1998, and a companion

section of NUREG-0800 to guide licensees on making risk-informed changes to plant technical

specifications. 'The agency uses this regulatory guide as well as Regulatory Guide 1.174 to

improve plant technical specifications. The industry and the NRC continue to increase the use

of PRA in developing improvements.to technical specifications. The following summarizes the

major accomplishments in this area:

Initiative 1, "Modified End States": This initiative would allow (following a risk assessment)

some equipment to be repaired during hot shutdown rather than cold shutdown. The

topical reports supporting this initiative for both boiling water reactor and Babcock &

Wilcox plants have been approved. The Westinghouse topical report, submitted

September 2005, is under review.

Initiative 4b, "Risk-Informed Completion Times": The overall objective of this initiative is

to modify technical specifications to reflect a configuration risk management approach

that is more consistent with'the approach of the maintenance rule (10 CFR 50.65(a)(4)).

Draft industry guidance and pilot plant applications are undergoing review. The South

Texas Project review will be completed in FY 2007. The Fort Calhoun Station review will

be completed in FY 2008.

Initiative 5b, "Risk informed method for Control of Surveillance Frequencies": This

initiative would allow licensees to modify the frequency of technical specification

surveillances based on test and risk data. The staff approved the use of this initiative for

Limerick Generating Station as a pilot plant in September 2006. The staff is currently

reviewing industry generic guidance that may be adopted by numerous licensees in

beginning in the summer of 2008.

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Initiative 6, "Modification of Limiting Condition for Operation 3.0.3, "Actions and

Completion Times": The industry is in the process of resolving discrepancies between its

Combustion Engineering topical report and the NRC's Topical Report Safety Evaluation.

The revised documents are expected to be resubmitted to the NRC in middle of 2008. A

BWR topical report is projected to be submitted by September 2007.

Initiative 7, "Non-Technical Specifications Support System Impact in Technical

Specifications System Operability": This initiative would permit a risk-informed delay time

before entering Limiting Condition for Operation actions for inoperability attributable to a

loss of support function provided by equipment. Guidance documents have been

approved for snubbers and hazard barriers and the industry is preparing additional

proposals.

Initiative 8, "Remove/Relocate non-safety and non-risk significant systems from

Technical Specifications": This initiative would review technical specifications to remove

certain system functions that had been included solely because they were judged to risk

significant at one time but could be shown by additional analysis not to be. The industry

and staff are in preliminary discussions on this initiative.

10.3.4 Development of Standards

The NRC worked with ASME to develop a national consensus standard for PRA related to

internal initiating events. ASME issued the standard, ASME-RA-5-2002, in April 2002. ASME

published the first two addenda to the standard in December 2003 and December 2005.

In parallel, the staff worked with the American Nuclear Society (ANS) to develop companion

standards covering PRAs for external events, fire, low power, and shutdown operations. Work

on these standards is progressing.

In January 2007, the NRC published Revision 1 to Regulatory Guide 1.200, "An Approach for

Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-informed

Activities." This revision incorporated lessons learned from licensee-piloted risk-informed

applications that rely on the regulatory guide. The agency plans further revisions to the

regulatory guide to incorporate the ANS standards as they are published.

10.4 Safety Culture

An important means to implement any policy that gives due priority to safety is to foster a strong

safety culture in the organization. The following discussion focuses upon safety culture, and

efforts to improve safety culture, in the NRC and in the nuclear industry.

10.4.1 NRC Monitoring of Licensee Safety Culture

This section covers the policies, programs, and practices that apply to licensee safety culture.

10.4.2.1 Background

Section 6.3.2 of this report describes the Reactor Oversight Process. Based on lessons learned

from the Davis-Besse reactor pressure vessel head degradation event, and other

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considerations, the NRC enhanced the Reactor Oversight Process to more fully address safety

culture to identify safety culture problems earlier so that corrective steps can be taken to

address the problems and prevent further plant performance degradation.

10:4.2.2 Enhanced Reactor Oversight Process

The NRC has adopted the IAEA International Nuclear Safety Advisory Group's definition of

safety-culture provided in Safety Series No.75-INSAG-4, "Safety Culture," issued February 1991,

as "that assembly of characteristics and attitudes in organizations and individuals which

establishes that, as an overriding priority, nuclear safety issues receive the attention warranted

by their significance."

On the basis of a review of safety culture attributes developed and/or applied by the IAEA, the

Nuclear Energy Agency, INPO, regulatory bodies in other countries, other domestic

organizations, staff expertise, and input and feedback from NRC stakeholders, the staff

identified the following components as being important to safety culture:

* decisionmaking

* resources

* work control

* work practices

* corrective action program

* operating experience

* self- and independent assessments

* environment for raising safety concerns

* preventing, detecting, and mitigating perceptions of retaliation

* accountability

* continuous learning environment

* organizational change management

* safety policies

Each one of the safety culture components is defined in a greater level of detail (e.g., crosscutting aspects) within the Reactor Oversight Process inspection guidance documents. The

Reactor Oversight Process applies the safety culture components, and their associated aspects,

in different ways. The first nine safety culture components are applied in the baseline inspection

and assessment program. All 13 safety culture components are applied in selected baseline,

event followup, and supplemental inspection procedures.

Licensees are performing periodic voluntary self-assessments of safety culture in accordance

with industry guidelines. There are no regulatory requirements for licensees to perform safety

culture assessments routinely. However, depending on the extent of deterioration of licensee

performance, the NRC has a range of expectations regarding regulatory actions and licensee

safety culture assessments, as described below.

The Reactor Oversight Process employs a graded approach in which plants that are performing

in a specified manner warrant only a routine level of inspection and oversight. However, as

licensee performance deteriorates, inspection and oversight become increasingly more intrusive

to ensure safe plant operation. The Reactor Oversight Process safety culture enhancements

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continue to allow licensees to self-diagnose and implement corrective actions for their

performance problems before the NRC performs followup inspections.

For most licensees (i.e., those listed in the Licensee Response column of the Reactor Oversight

Process Action Matrix), the NRC performs the baseline inspection program. In the routine or

baseline inspection program, the inspector will develop an inspection finding and then identify

whether an aspect of a safety culture component is a significant causal factor of the finding. The

inspection findings are communicated to the licensee along with the associated safety culture

aspect.

The NRC revised the inspection procedure that focuses on problem- identification and resolution

to allow inspectors to have the option to review licensee self-assessments of safety culture, The

problem identification and resolution inspection procedure also instructs inspectors to be aware

of safety culture components when selecting samples. In addition, questions related to safetyconscious work environment were enhanced in the procedure.

The agency revised the event response inspection procedures to direct inspection teams to

consider contributing causes related to the safety culture components as part of their efforts to

fully understand the circumstances surrounding an event and its probable cause(s).

As part of the assessment process (conducted twice per year), the NRC considers the aspects

of safety culture components associated with inspection findings to determine whether common

themes exist at a plant. If, over three consecutive assessment periods (i.e., 18 months), a

licensee has the same safety culture issue with the same common theme, the NRC may ask the.

licensee to conduct a safety culture self-assessment.

As licensee performance declines (Regulatory Response-column of the Reactor Oversight

Process Action Matrix), the inspectors, through a specific supplemental inspection procedure,

verify that the licensee's root cause, extent of condition, and extent-of cause evaluations for the

risk-significant finding(s) appropriately considered the safety culture components.

When the licensee performance degrades further (Degraded Cornerstone column of the Reactor

Oversight Process Action Matrix), the NRC expects that the licensee's root cause evaluation for

the risk-significant finding(s) determined whether any safety culture component contributed to

the risk-significant performance issues. If through the conduct of a specific supplemental

inspection procedure, the NRC determines that the licensee did not recognize that safety culture

components caused orsi'gnificantly contributed to the risk-significant performance issues, the

NRC may request the licensee to complete an independent assessment of its safety culture.

Finally, for licensees with more significant performance degradation (Multiple/Degraded

Cornerstone column of the Reactor Oversight Process Action Matrix), the NRC will expect the

licensee to conduct a third-party independerit assessment of its safety culture. The NRC will

review the licensee's assessment and will conduct an independent assessment of the licensee's

safety culture via a specific supplemental inspection procedure that was substantiallyrevised to

provide guidance for these assessments. The staff will be applying this revised inspection

procedure for the first time at the Palo Verde plant in the latter part of 2007.

In July 2006, the NRC implemented revisions to the Reactor Oversight Process inspection and

assessment processes related to safety culture. The NRC inspectors were trained on safety

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culture in general and on the changes to the Reactor Oversight Process before implementation.

Ongoing inspector training now includes safety culture topics. The NRC is reviewing the

changes to the Reactor Oversight Process over an 18-month initial implementation period that

began in July 2006 and will report on the implementation lessons learned in 2008.

The safety culture changes made to the Reactor Oversight Process were intended to provide the

NRC staff with (1) better opportunities to consider safety culture weaknesses and to encourage

licensees to take appropriate actions before significant performance degradation occurs, (2) a

process to determine the need to specifically evaluate a licensee's safety culture after

performance problems have resulted in the placement of a licensee in the Degraded

Cornerstone column of the Reactor Oversight Process Action Matrix, and (3) a structured

process to evaluate the licensee's safety culture assessment and to independently conduct a

safety culture assessment for a licensee in the Multiple/Repetitive Degraded Cornerstone

column of the action matrix.

By using the existing Reactor Oversight Process framework, the NRC's safety culture oversight

activities are based on a graded approach and remain transparent, understandable, objective,

risk informed, performance based, and predictable.

10.4.2 The NRC Safety Culture

The NRC has programs to promote an open, collaborative working environment that encourages

all employees and contractors to promptly voice differing .views without fear of retaliation.

Additionally, the agency assesses the effectiveness of these programs and the working

environment through the triennial Safety Culture and Climate Survey.

In fall 2005, the NRC's Inspector General (IG), with the assistance of a contractor research firm,

conducted the Safety Culture and Climate Survey (see OIG-06-A-08, "OIG 2005 Survey of

NRC's Safety Culture and Climate," February 10, 2006). The agency conducted similar surveys

in 2002 and 1998. All NRC employees and managers were offered the opportunity to take the

survey, and 70 percent responded, as compared to 53 percent in 2002. The evaluation report of

the survey results is on the NRC's public Web site.

The purposes of the 2005 Safety Culture and Climate Survey were to (1) gauge the NRC's

safety culture and climate, (2) compare results against the 2002 survey, and (3) compare results

to Government and national benchmarks. This triennial assessment allows the NRC to address

identified areas for improvement and maintain or improve areas of strength.. Survey questions

were grouped into 18, categories representing the major topics of the NRC's safety culture and

climate. The categories singled out for particular comment are summarized below.

Compared to other benchmarks, the 2005 NRC survey results are statistically more favorable in

12 of 16 categories compared to the U.S. national norm and in 14 of 16 categories compared to

the U.S. research and development norm. No categories in the survey results when compared

to either of the norms are statistically less favorable. Comparing the 2005 to the 2002 NRC

results, 16 of 18 categories have statistically improved, from 13 points in Communication to 3

points in Clarity of Responsibilities. The remaining two of 18 categories did not experience a

statistically significant change from the 2002 results.

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Compared to the 2002 survey, the NRC improved in essentially all areas, with the largest gains

realized in communication, mission and strategic planning, employee engagement, recruiting,

developing and retaining staff, and management leadership. Areas for improvement were

workload and stress, knowledge transfer, and the'use of the Differing Professional Opinion

Program to raise concerns.

The NRC is acting to address the survey responses. Examples of the agency's responses to

two specific areas for improvement, knowledge transfer and differing professional opinions,

follow. The agency's Knowledge Management Initiative, inaugurated soon after the 2005

survey, provides a cross-office effort to capture and transfer knowledge through a variety of

tools, such as formal and informal training, databases, electronic reading rooms, interviews,

procedures, desk references, communities of practice, and Web sites. The NRC considers KM

to be part of the strategic management of human capital, along with strategic workforce

planning, recruitment, and training and development. SECY-06-0164, "NRC Knowledge

Management Program," dated July 24, 2006, discusses this issue further.

The NRC's Differing Professional Opinion Program, as defined in Management Directive 10.159,

"The NRC Differing Professional Opinions Program," dated May 16, 2004, allows NRC

employees to bring safety concerns or other issues important to the agency's mission to senior

NRC management and, where appropriate, to the Commission. This program supports the

agency policy to maintain a working environment that encourages employees to make known

their best professional judgments even though they may differ from the prevailing staff view,

disagree with a management decision or policy position, or take issue with a proposed or

established agency practice involving technical, legal, or policy issues. The program ensures

the full consideration and timely disposition of differing professional opinions by affording an

independent, impartial review by knowledgeable personnel who review the concern to determine

the need for regulatory action. The program may be used without fear of retaliation, pressure,

penalty, or unauthorized divulgence.

On November 29, 2006, the Executive Director for Operations issued draft Management

Directive and Handbook 10.158, "NRC Non-Concurrence Process," as interim policy to provide

agencywide instructions and guidance on raising concerns during the document concurrence

process. One of this new program's objectives is to promote discussion and consideration of

differing views on documents in the concurrence process before the prevailing staff view is fully developed or management or policy decisions are made.

These programs, along with the NRC's open door policy, affirm the agency's commitment to,

and expectations for, an open, collaborative working environment. Employees are reminded of

their responsibility to raise concerns as early as possible, engage in discussions, and seek

solutions using one of the above approaches.

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ARTICLE 11. FINANCIAL AND HUMAN RESOURCES

1. Each Contracting Party shall take the appropriate steps to ensure that adequate

financial resources are available to support the safety of each nuclear installation

throughout its life.

2. Each Contracting Party shall take the appropriate steps to ensure that sufficient

numbers of qualified staff with appropriate education, training, and retraining are

available for all safety-related activities in or for each nuclear installation,

throughout its life.

This section explains the requirements regarding the financial resources that licensees must

have to support the nuclear installation throughout its life and the regulatory requirements for

qualifying, training, and retraining personnel.

The NRC updated this section to incorporate changes in the dollar amounts for liability under the

Price-Anderson Act and thoroughly revised section 11.2.

11.1 Financial Resources

Adequate funds for safe construction, operation, and decommissioning are necessary for the

protection of public health and safety. Although there does not appear to be a consistent

relationship between a licensee's finances and operational safety, some evidence suggests that

financial pressures have limited the resources devoted to corrective actions, plant

improvements, and other safety-related expenditures. Furthermore, because a power reactor

must operate to provide revenues for eventual plant decommissioning, any shutdown of a plant

before its owner has accumulated sufficient funds for decommissioning could potentially hinder

the safe decommissioning of that plant.

Additionally, many States (i.e., of the United States) have initiated or completed actions to

economically deregulate their nuclear power plants. Traditionally, nuclear power plant owners in

many States have been large, vertically integrated companies with substantial assets in

generation, transmission, and distribution. In exchange for having exclusive franchises to

provide electric power in defined geographical areas, nuclear plant owners have had the rates

they charge to their customers regulated by State governmental bodies. This system of ratebased regulation has ensured a source of funds for construction, operation, and

decommissioning of nuclear power plants. Nonetheless, this model of rate-based regulation has

been changing and the United States has adjusted its processes.

The NRC distinguishes between financial qualifications for construction and operation, as well as

decommissioning of nuclear power plants, and has separate regulations and programs that

apply to each. The NRC also implements programs to ensure that the public has financial

protection for bodily injury and property damage losses in the event of an accident. Finally, the

agency has implemented requirements to ensure that licensees have insurance to help pay

onsite recovery costs resulting from accidents and to provide funds for postaccident restart or

decommissioning.

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11.1.1 Financial Qualifications Program for Construction and Operations

This section explains the financial qualifications program for construction and operations. It

explains NRC reviews for construction permits, operating licenses, combined licenses,

postoperating nontransferred licenses, and license transfers.

Section 182.a of the Atomic Energy Act provides that "each application for a license ...shall

specifically state such information as the Commission, by rule or regulation, may determine to be

necessary to decide such of the technical and financial qualifications of the applicant.. .as the

Commission may deem appropriate for the license." To implement this provision, the NRC has

developed the regulations and guidance discussed below.

11. 1.1.1 Construction Permit Reviews

As required by 10 CFR 50.33(f)(1), applicants for construction permits must submit information

that "demonstrates that the applicant possesses or has reasonable assurance of obtaining the

funds necessary to cover estimated construction costs and related fuel cycle costs."

Appendix C, "A Guide for the Financial Data and Related Information Required to Establish

Financial Qualifications for Facility Construction Permits," to 10 CFR Part 50 provides more

specific directions for evaluating the financial qualifications of applicants.

11. 1.1.2 Operating License Reviews

An "electric utility" as defined in 10 CFR 50.2, "Definitions," is "any entity that generates or

distributes electricity and which recovers the cost of this electricity, either directly or indirectly,

through rates established by the entity itself or by a separate regulatory authority." Electric

utilities are exempt under 10 CFR 50.33(f) from reviews of financial qualifications of applications

for operating licenses. The reason for this exemption is that cost-of-service rate regulation,. as it

has existed in the United States, has ensured that ratepayers provide a source of funds-for the

safe operation of nuclear power plants. Applicants for operating licenses that are not electric

utilities are required under 10 CFR 50.33(f)(2) to submit information that demonstrates that they

possess or have reasonable assurance of obtaining the necessary funds to cover estimated

operating costs. Nonelectric-utility applicants for operating licenses are also required to submit

estimates for the total annual operating costs for each of the first 5 years of operation of their

facilities and must indicate the sources of funds to cover operating costs.

11.1.1.3 Combined License Application Reviews

As authorized in 10 CFR Part 52, applicants may apply for a combined construction permit and

operating license. Under 10 CFR 52.77, "Contents of Applications; Technical Information," such

applications must contain all of the information required under 10 CFR 50.33, including

information regarding financial qualifications. The procedures described above are used to

review future combined license applications that the NRC receives.

11.1.1.4 Postoperating License Nontransfer Reviews

The NRC does not systematically review the financial qualifications of power reactor licensees

once it has issued an operating license, other than for license transfers as described below...

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However, as provided in 10 CFR 50.33(f)(4), the NRC can seek additional information on

licensees' financial resources if the agency considers such information appropriate.

11.1.1.5 Reviews of License Transfers

The NRC regulations in10 CFR 50.80 require agency review and approval of transfers of

operating licenses, including licenses for nuclear power plants that are owned or operated by

electric utilities. The NRC performs these reviews to determine whether a proposed transferee

or new owner is technically and financially qualified to hold the license.

NUREG-1 577, Revision 1, "Standard Review Plan on Power Reactor Licensee Financial

Qualifications and Decommissioning Funding Assurance" describes the agency's overall review

process for applicant and licensees' financial qualifications for nuclear power plant construction

and operation.

11.1.2 Financial Qualifications Program for Decommissioning

Among~other sections of the Atomic Energy Act, Section 182.a establishes the basis for NRC's

decommissioning funding assurance regulations and guidance; the previous U.S. National

Report listed specific regulations and guidance.

11.1.3 Financial Protection Program for Liability Claims Arising from Accidents

The Price-Anderson Act, enacted in 1957, became Section 170 of the Atomic Energy Act and

governs the U.S. financial protection program. Section 170 (with related definitions in

Section 11) provides the financial and-the legal framework to compensate those who suffer

bodily injury or property damage as a result of accidents at nuclear facilities covered by the law's'

provisions. The NRC regulations implementing the provisions of Section .170 for NRC licensees

are codified in 10, CFR Part 140, "Financial Protection Requirements and Indemnity

Agreements."

The Price-Anderson Act was enacted to (1) remove the deterrent to private sector participation

in atomic energy presented by the threat of potentially enormous liability claims in the event of a

catastrophic nuclear accident and (2) ensure that adequate funds are available to the public to

satisfy liability claims if such an accident were to occur.

The Price-Anderson Act was revised most recently in 2005, when Congress renewed the

Commission's authority to cover new facilities until 2020. Under current law, power reactors

over 100 megawatts electric must contribute to a pool that replaces the Government as the

second provider of funds if the first layer of financial protection (liability insurance-now

$300 million) is exhausted.

The reactor operators are required after an accident to pay into a "retrospective premium pool,"

in maximum annual installments not to exceed $15 million, up to a total of $95.8 million each.

But payment is called for only if the accident exhausts the first layer of financial protection, and

only if, and to the extent that, additional funds are needed to pay the damages. With 104

reactors currently participating in the system, the total financial protection available under the

Price-Anderson Act for any one accident is $10 billion ($300 million primary coverage +

(95.8 million per reactor x 104 reactors)); $10 billion is also the limit on liability. As reactors

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leave the retrospective premium system as a result of permanent closure or join as the result of

construction of new reactors, this coverage limit may fall or rise. A change in the limit may also

occur when the $95.8 million contribution is adjusted for inflation, as must be done every

5 years. In any event, Congress will address any damages exceeding the total sum that

reactors must contribute to the pool and will decide upon the next steps needed for

compensation.

The public is significantly benefitted by another feature of the act. Claimants need only prove

that the accident caused their injury in. order to receive compensation for damages from any

accident witlh significant offsite releases of radiation (i.e., an "extraordinary nuclear occurrence").

Neither proof of fault nor proof of what caused the accident is necessary.

Claims for more than 150 alleged incidents involving nuclear material have been filed under

various liability policies since the inception of the Price-Anderson Act in 1957. The insured

losses and expenses paid so far total more than $125 million. Of this amount, most payments

arose out of the accident at TMI Unit 2.

11.1.4 Insurance Program for Onsite Property Damages Arising from Accidents

Among other sections of the Atomic Energy Act, Section 182.a provides the basis for the NRC's

onsite property damage insurance requirements for operating nuclear power reactors contained

in 10 CFR 50.54(w).

The U.S. nuclear industry has not experienced an accident involving significant radioactive

contamination since TMI Unit 2.

11.2 Regulatory Requirements for Qualifying, Training, and Retraining Personnel

This section explains the regulatory requirements for'qualifying, training, and retraining

personnel. It discusses the governing documents, the process for implementing requirements,

and experience. It also discusses the INPO accreditation activities.

11.2.1 Governing Documents and Process

The training requirements for licensed operators and licensed senior operators are regulated by

10 CFR Part 55, "Operators' Licenses," which allows facility licensees to have operator

requalification program content that is derived using a systems approach to training (SAT; as

defined in 10 CFR 55.4) or that meets the requirements outlined in 10 CFR 55.59(c). Subpart D,

"Applications," of 10 CFR Part 55 requires that operator license applications must contain

information about an individual's training and experience, unless the facility licensee certifies that

the applicant has successfully completed a Commission-approved training program that is SATbased and uses an acceptable simulation facility.

The operator licensing process at power reactors includes a generic fundamentals examination

covering the theoretical knowledge that is required to operate a nuclear power plant; license

applicants must pass the generic fundamentals examination before they can take a site-specific

examination. The site-specific examination consists of a written examination and an operating

test that includes a plant walkthrough and a dynamic performance demonstration on a simulation

facility.

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The NRC staff has transferred most of the responsibility for developing site-specific licensing

examinations to facility licensees. In 1999, the NRC amended 10 CFR Part 55 to allow nuclear

power reactor licensees to prepare the written examinations and operating tests that the agency

uses to evaluate the competence of applicants for operators' licenses at those facilities.

Licensees that elect to prepare their own examinations are required to establish procedures to

control examination security and integrity. The licensees prepare and submit proposed

examinations and operating tests to the NRC according to the guidance in NUREG-1021,

"Operator Licensing Examination Standards for Power Reactors," Revision 9, issued July 2004.

The NRC reviews the facility-prepared examinations, prepares examinations for facility licensees

upon request, administers all operating tests, makes the final licensing decisions, and issues the

licenses.

As required by 10 CFR 50.120, "Training and Qualification of Nuclear Power Plant Personnel,"

training programs must be established, implemented, and maintained using a SAT approach for

eight categories of nonlicensed workers at nuclear power plants and the shift supervisor, who is

licensed in accordance with 10 CFR Part 55. These provisions complement the requirements

for training based on a systems approach for the requalification of licensed operators and

licensed senior operators. Regulatory Guide 1.8, "Qualification and Training of Personnel for

Nuclear Power Plants," Revision 3, issued May 2000, contains guidance to implement the

regulations.

The NRC continues to endorse the training accreditation process managed by INPO. The staff

recognizes that training programs developed in accordance with INPO guidelines and accredited

by the National Nuclear Accrediting Board (NNAB) are SAT-based; therefore, accredited

programs are considered to be consistent with the regulations in 10 CFR Part 55 and 10 CFR

Part 50.120. The NRC also recognizes that INPO-managed accreditation and associated

training evaluation activities are an acceptable means of self-improvement in training. Such

recognition encourages industry initiative and reduces NRC evaluation and inspection activities.

In accordance with its memorandum of agreement with INPO, the NRC monitors INPO

accreditation activities as part of its continuing assessment of the effectiveness of the industry's

training programs. Specifically, the NRC staff observes selected accreditation team visits and

NRC managers periodically observe NNAB meetings. These observations are intended to

monitor the implementation of programmatic aspects of the accreditation process, but they also

provide an opportunity to assess the selected performance areas of facility licensees.

If the NNAB has concerns regarding the performance of an accredited training program, it will

place the program on probation. This does not necessarily place a training program in

noncompliance with either 10 CFR Part 55 or 10 CFR 50.120, since training programs are

accredited to a standard of excellence rather than a minimum level of regulatory compliance.

However, the NRC does review the circumstances leading to the probation to ensure safe

operations and continued compliance with the regulations.

The NNAB may also withdraw accreditation in response to major deficiencies in a licensee's

accredited training program. If accreditation is withdrawn, the NRC would request that the

licensee report the circumstances of the withdrawal for the staff to determine the significance of

the issues related to the withdrawal. If the NRC determines that compliance with the regulations

is not affected, it may not be necessary to take any further action. If the withdrawal is linked to a

breakdown in the training process or a safety-significant issue, the NRC will conduct an

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immediate inspection focused on the process problem or'safety issue(s). The agency would

take further action, such as issuing confirmatory action letters or orders, if appropriate.

The NRC monitors industry performance in implementing the training requirements of

10 CFR Part 50 and 10 CFR Part 55 by (1) reviewing licensee event reports and inspection

reports for training issues, (2) observing the accreditation process, and (3) reviewing the results'

of operator licensing activities. Guidance for periodically inspecting the licensed operator

requalification training program at every facility is given in Inspection Procedure (IP) 71111.11,

"Licensed Operator Requalification Program." In addition, when appropriate for cause, the NRC

will use IP 41500, "Training and Qualification Effectiveness," which references the guidance in

NUREG-1220, "Training Review Criteria and Procedures," Revision 1, issued January 1993, to

verify compliance with SAT requirements.

11.2.2 Experience

The NRC reviewed training issues contained in licensee event reports and inspection reports

during 2006 using data from the Human Factors Information System, which is described in

Article 12. The review revealed that the proportion of human performance issues attributable to

training, for U.S. nuclear power plants, decreased from eight percent in 1999 to four percent in

2005. The training-related issues identified were concentrated in two subcategories, "training

less than adequate" and "individual knowledge less than adequate." The NRC annually

assesses the effectiveness of training in the nuclear industry and prepares a report of its

findings; the reports for 1999 through 2005 are available on the NRC's public Web site.

Although the NRC identified some limited specific weaknesses in training programs, all

indicators suggest that the industry is successfully implementing training programs in

accordance with the regulations.. The NRC will continue to monitor selected performance areas,

emphasizing the identification and resolution of training process problems.

An example of this monitoring process is the for-cause training inspection conducted at Virgil C.

Summer Nuclear Station in May 2006. The NRC performed this inspection in response to the

high failure rate on the written portion of the initial operator license examination administered in

January 2006. The inspection provided reasonable assurance that the operator license

applicants had successfully completed the initial training program and that the program meets

SAT requirements.

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ARTICLE 12. HUMAN FACTORS

Each Contracting Party shall take the appropriate step to ensure that the capabilities and

limitations of human performance are taken into account throughout the life of a nuclear

installation.

This section explains the NRC program on human performance. The seven major areas under

this program are (1) human factors engineering issues, (2) emergency operating procedures and

plant procedures, (3) working hours and staffing, (4) fitness-for-duty, (5) human factors

information system, (6) support to event investigations and for-cause inspections, and (7)

training.

The NRC updated this section to incorporate new documents, issues, and experience.

12.1 Goals and Mission of the Program

The NRC has a comprehensive program for ensuring that human performance is properly

addressed in a risk-informed regulatory framework for maintaining'reactor safety. The NRC

developed the program based on reviewing risk information and activities in the domestic and

international nuclear industry.

12.2 Program Elements

The Reactor Oversight Process (discussed in Article 6) focuses on cornerstones of safety that

are assessed through a combination of performance indicators and risk-informed inspections.

The inspections focus on risk-significant activities and systems related to the cornerstones. The

three elements that are considered cross-cutting with respect to the cornerstones are human

performance, safety-conscious work environment, and corrective actions. The Human

Performance Program has contributed directly to the development of a supplemental inspection

procedure related to the human performance cross-cutting element. The Human Performance

Program is also engaged in the other two elements, since a safety-conscious work environment

and many of the actions involved in corrective action programs result from human performance

problems.

The Human Performance Program also supports the Risk-Informed and Performance-Based

Plan by generating, collecting, and evaluating data on human performance for use in human

reliability analysis models. The staff evaluates information to gain insights to support

risk-informed regulation and to provide human performance data for human reliability analysis.

The Human Performance Program monitors technological developments and emerging issues in

an effort to prepare the NRC for the future. Two ongoing activities include developing regulatory

guidance for reviewing designs of control stations and processing requests related to

deregulation. Licensees are replacing aging analog controls and displays with digital

components, and the NRC must be prepared to review safety issues with respect to

human-system interfaces resulting from such new designs and technologies. With regard to

deregulation, the NRC has been processing numerous industry requests to transfer operating

licenses, which may involve changes in organizational structure affecting human performance.

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12.3 Significant Regulatory Activities

The NRC performs significant regulatory activities in the following seven areas to address

human performance:

(1) human factors engineering Issues

(2) emergency operating procedures and plant procedures

(3) shift staffing

(4) fitness-for-duty

(5) human factors information system

(6) support to event investigations and for-cause inspections

(7) training

The first six activities are described below; Article 11 describes training.

12.3.1 Human Factors Engineering Issues

This section discusses human factors activities related to engineering issues, covering the

governing documents and process to carry out requirements and experience.

Governing Documents and Process. The NRC evaluates the human factors engineering design

of the main control room and control centers outside of the main control room using Chapter 18

of NUREG-0800, NUREG-0700, "Human System Interface Design Review Guideline,"

Revision 2, issued May 2002, and NUREG-071 1, "Human Factors Engineering Program Review

Model, Revision 2" issued February 2004. These documents provide guidance for the review of

human-system interface issues in connection with the design certification of nuclear installations

and the NRC's inspection program. The NRC also uses NUREG-1 764, "Guidance for the

Review of Changes to Human Actions," to review license amendment requests that credit the

use of manual actions. Moreover, IN 97-78, "Crediting of Operator Actions in Place of Automatic

Actions and Modifications of Operator Actions, Including Response Times," dated October 23,

1997, identifies references that the NRC uses to review the completion times of operator manual

actions and how the actions will be reflected in the licensee's emergency procedures and

operator training.

Experience. The NRC reviews licensees' requests that involve aspects of human factors

engineering. Examples include crediting operator manual actions in amendments to plant

technical specifications, transferring facility operating licenses, and increasing the reactor's

authorized power level (i.e., power uprates). Recent license amendment requests from Fort

Calhoun Station and R.E. Ginna Nuclear Power Plant are examples of NRC reviews involving

new or modified operator manual actions. The amendments from Fort Calhoun proposed

crediting manual actions to isolate steam generator blowdown within a certain time during a loss

of main feedwater event and to reduce the number of containment spray pumps needed in

response to a LOCA. The amendment request from Ginna also involved operator manual

actions to realign safety injection pumps for recirculation following a LOCA.

The NRC has also evaluated a number of requests to transfer facility operating licenses, which

affected management and organization, staffing, and technical qualifications. The NRC used

Chapter 13 of NUREG-0800 as the principal guidance for these reviews.

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The NRC also reviews and approves requests for power uprates from currently licensed plants.

For such requests, the NRC examines the effect of the power uprate on plant procedures,

controls, displays, and alarms, and required operator actions using.Review Standard (RS-001),

"Review Standard for Extended Power Uprates," Section 2.11.1 (in both BWR and PWR

sections). The NRC issued this guidance in December 2003; it can be found on the NRC's'

public Web site along with additional general information on power uprates. The agency

recently reviewed and approved power plant uprates for R.E. Ginna, Beaver Valley Units 1 and

2, and Vermont Yankee Nuclear Power Station.

12.3.2 Emergency Operating Procedures and Plant Procedures

Licensees must have programs to develop, implement, and maintain emergency operating and

plant procedures. (Article 16 discusses emergency preparedness; this discussion is limited to

the human factors aspect of emergency operating procedures.)

Governing Documents and Process. On December 17, 1982, 'the NRC issued GL 82-33,

"Requirements for Emergency Response Capability," which transmitted Supplement 1 to

NUREG-0737, "Requirements for Emergency Response Capability," requires each licensee to

submit a set of documents for developing emergency operating procedures.

Experience. No significant examples related to emergency operating and plant procedures were

identified since 2004.

12.3.3 Shift Staffing

Governing Documents and Process. Paragraph (m) of 10 CFR 50.54, "Conditions of Licenses,"

specifies the minimum number of licensed operators and senior operators that are required for

nuclear power reactor facilities. Appendix R, "Fire Protection Programs for Nuclear Power

Facilities Operating Prior to January 1, 1979," and Appendix E, "Emergency Planning and

Preparedness for Protection and Utilization Facilities," to 10 CFR Part 50 contain the NRC

staffing requirements for fire brigades and emergency response personnel.

In September 2002, the NRC began work on a process to evaluate exemption requests from the

requirements in 10 CFR 50.54(m) resulting from the changing demands and new technologies

presented by advanced reactor control room designs and significant light-water reactor control

room upgrades. In July 2005, the NRC published NUREG-1 791, "Guidance for Assessing

Exemption Requests from the Nuclear Power Plant Licensed Operator Staffing Requirements

Specified in 10 CFR 50.54(m)." The purpose for reviewing the exemption requests is to ensure

public health and safety by verifying that the applicant's staffing plan and supporting analyses

sufficiently justify the requested exemption. NUREG/CR-6838, "Technical Basis for Assessing

Exemptions from Nuclear Power Plant Licensed Operator Staffing Requirements

10 CFR 50.54(m)," issued February 2004, explains the justification for the recommended

process.

Experience. No significant examples related to shift staffing were identified for 2004-2006.

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12.3.4 Fitness-for-Duty

This section discusses the NRC's requirements pertaining to the fitness-for-duty of nuclear

power plant workers, including requirements regarding the control of work hours and

management of worker fatigue.

GoverninQ Documents and Process. As required by 10 CFR Part 26, "Fitness for Duty

Programs," each licensee authorized to operate or construct a nuclear power reactor must

implement a fitness-for-duty program for all personnel having unescorted access to the

protected area of its plant. As performance objectives, 10 CFR Part 26 requires that licensees

establish programs that (1) provide reasonable assurance that nuclear power plant personnel

perform their tasks in a reliable and trustworthy manner and are not under the influence of any

substance, legal or illegal, or mentally or physically impaired from any cause, (2) provide

reasonable measures for the early detection of persons who are not fit to perform activities, and

(3) have a goal of achieving a-drug-free workplace and a workplace free of the effects of such

substances.

The NRC issues annual reports on statistical data and lessons learned by licensees from their

fitness-for-duty program performance reports. The most recent of these is NRC Information

Notice 2006-30, "Summary of Fitness-for-Duty Program Performance Reports for Calendar

Years 2004 and 2005," dated December 21, 2006. The NRC is currently completing a similar

report that will provide data for 2006. A project to automate the reporting and trending of

performance data using a Web-based approach is ongoing. In addition, the NRC has

established an email address for licensees and individuals to submit fitness-for-duty questions,

as well as a Web site where performance reports and the answers to frequently asked questions

are publicly available.

Regarding worker fatigue, in 1982, the NRC issued its."Policy on Factors Causing Fatigue of

Operating Personnel at Nuclear Reactors." The objective of the policy is for licensees to ensure,

to the extent practicable, that personnel who .perform safety-related functions are not assigned

to shift duties while in a fatigued condition that can significantly reduce their mental alertness or

decisionmaking ability. Licensees have implemented the policy through reactor facility technical

specifications that require administrative control of personnel work hours. In April 2003, the

NRC issued Order-03-038, "Order for Fitness-for-Duty Enhancements for Nuclear Security Force

Personnel," requiring licensees to implement compensatory measures to control the work hours

of security personnel at nuclear power plants. The order requires licensees to maintain the work

hours of security force personnel below specified levels, monitor for individual fatigue, and

establish a process to follow if an individual declares that he or she is unfit for duty as a result of

fatigue. The NRC issued the order in response to concerns associated with a significant

increase in the work hours of security personnel at U.S. nuclear facilities following the terrorist

attacks on September 1-1, 2001.

Experience. When 10 CFR Part 26 was published in 1989, the Commission directed the NRC

staff to continue to analyze licensee programs, assess the effectiveness of the rule, and

recommend appropriate improvements or changes. In 2006, following years of stakeholder

interactions, the NRC staff proposed a comprehensive amendment of the rule to update and

enhance the fitness-for-duty requirements. The proposed amendment, which was provided to

the Commission via SECY-06-0244, "Final Rulemaking-10 CFR Part 26-Fitness-for-Duty

Programs," on December 22, 2006, includes provisions for specimen validity testing, stronger

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sanctions for refusing to submit to testing and attempts to circumvent the testing process, and

increased legal protections for those tested. The proposed amendment would also establish

requirements for the management of worker fatigue, including requirements for licensees to (1)

train workers in methods for managing fatigue, (2) establish procedures for appropriately

managing instances of individuals declaring they are unfit for duty because of fatigue, (3) limit

the work hours of individuals who perform functions important to the protection of public health

and safety, and (4) submit an annual report that describes how often the licensee authorizes

individuals to exceed the work hour limits. Full implementation of the rule is expected to take at

least one year from the date of publication in the Federal Register.

12.3.5 Human Factors Information System,

Governing Documents and Process. The Human Factors Information System is designed to

store, retrieve, sort, and analyze human performance information extracted from NRC inspection

and licensee event reports. Initiated in 1990, this automated information management system

can generate a variety of specialized reports that are not readily available from other NRC

sources. In 2006, improvements were made to this system to better align the coding scheme

with the Reactor Oversight Process (described in Article 6) and to enhance the system's search

capabilities. The Human Factors Information System now captures information related to

training, procedures and reference documents, fitness-for-duty, oversight, problem identification

I and resolution, communications, human-system interface and environment, and work planning

and practices.

Experience. The NRC responds to stakeholder and public inquiries and data requests on this

system on a regular basis. For example, inspectors use the data generated by this system in

preparing inspection activities related to human performance. In addition, the NRC's Office of

Research uses the data to support activities in human performance and human reliability

analysis. Other NRC program offices use the data to gain insights about human performance

and to monitor the frequency of human performance issues and to inform several types of

reports, such as internal operating experience reports and the NRC's annual report on the

effectiveness of training in the nuclear industry (discussed in Section 11.2.2 of this report). The

NRC also uses aWeb page to disseminate information on human performance issues at

individual nuclear power plant sites.

12.3.6 Support to Event Investigations and For-Cause Inspections and Training

Governing Documents and Process. NRC staff members with human factors expertise are often

included in special inspections, incident investigation team inspections, augmented team

inspections, event investigations, and supplemental inspections. Human factors experts have

assessed management effectiveness, procedures, training issues, staffing issues, humanmachine interfaces, personnel performance issues, safety-conscious work environment, and

safety culture.

For training issues, inspectors use IP 41500. For procedure issues, inspectors use IP 42001,

"Emergency Operating Procedures," and IP 42700, "Plant Procedures." For baseline

inspections under the Reactor Oversight Process (described in Article 6), inspectors use IP

71152, "Identification and Resolution of Problems," which is intended to establish confidence

that each licensee is, detecting and correcting problems in a manner that limits the risk to the

public and includes a review of the licensee's safety-conscious work environment. A key

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premise of the Reactor Oversight Process is that weaknesses in problem identification and

resolution programs will manifest themselves as performance issues that can be identified

during the baseline inspection program or by crossing predetermined indicator thresholds.

For supplemental inspections, IP 95003, "Supplemental Inspection for Repetitive Degraded

Cornerstones, Multiple Degraded Cornerstones, Multiple Yellow Inputs, or One Red Input," as

revised in October 2006, includes requirements for the NRC staff to review the licensee's thirdparty safety culture assessment and independently assess the licensee's safety culture. Staff

with technical expertise in human factors and safety culture will perform the safety culture

activities. The first implementation of revised IP 95003 is scheduled for the Fall 2007.

Experience.. In 2005 'the NRC staff with human factors expertise participated in an IP 95003

inspection at the Perry Nuclear Power Plant to assess human performance at the site. The

inspectors determined that a number of findings related to procedure adherence had strong

human performance contributions. The NRC returned to Perry in 2006 to assess the

effectiveness of the licensee's corrective actions in addressing the human performance issues

identified during the previous inspection. The evaluation included observing the use of human

performance error prevention tools during work activities in the plant, observing supervisors' use

and reinforcement of these tools, and reviewing the licensee's performance indicators in the

human performance area. Based on these activities, the inspectors concluded that the

licensee's corrective actions were effective in improving human performance. The inspectors

found that the licensee had adequately implemented commitments to address issues in the

human performance area and evaluated the effectiveness of the corrective actions.

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ARTICLE 13. QUALITY ASSURANCE

Each Contracting Party shall take the appropriate steps to ensure that quality assurance

programmes are established and implemented with a view to providing confidence that

specified requirements for all activities important to nuclear safety are satisfied

throughout the life of a nuclear installation.

This section explains QA policy and requirements and guidance for design and construction,

operational activities, and staff licensing reviews. It also describes QA programs and regulatory

guidance. This section was updated to discuss a new construction program for expected new

reactors.

13.1 Background

Nuclear power facilities must be designed, constructed, and operated in a manner that ensures

(1) the prevention of accidents that could cause undue risk to the health and safety of the public

and (2) the mitigation of adverse consequences of such accidents if they should occur. A

primary means for achieving these objectives is by establishing and effectively implementing a

nuclear QA program. Although a licensee may delegate aspects of the establishment or

execution of the QA program to others, the licensee remains ultimately responsible for the

program's overall effectiveness. Licensees perform a variety of self-assessments to validate the

effectiveness of their QA program implementation. The NRC reviews descriptions of QA

programs and performs onsite inspections to verify aspects of the program implementation.

13.2 Regulatory Policy and Requirements

Each applicant for a construction permit for a nuclear power plant is required by

10 CFR 50.34(a)(7) to describe its QA program in its preliminary safety analysis report. Also,

each applicant for a combined license is required by 10 CFR 52.79(b) to describe its QA

program in its preliminary safety analysis report. This program applies to the design, fabrication,

construction, and testing of safety-related plant equipment. Each applicant for a license to

operate a nuclear power plant is required by 10 CFR 50.34(b)(6)(ii) to provide a final safety

analysis report that details its managerial and administrative controls to ensure safe operation.

In both the preliminary and final safety analysis reports, the applicant must describe how it will

satisfy the applicable requirements of Appendix B, "Quality Assurance Criteria for Nuclear Power

Plants and Fuel Reprocessing Plants," to 10 CFR Part 50.

If a licensee wants to make changes in its QA program, it is required by 10 CFR 50.54(a)(3) to

inform the NRC of the changes. A licensee can make changes without prior NRC approval if the

.changes do not reduce the commitments in the program description as accepted by the NRC.

Nuclear QA criteria apply to all activities that affect the safety-related functions of SSCs that

prevent or mitigate the consequences of postulated accidents that could cause undue risk to the

health and safety of the public. High-level criteria for determining which plant SSCs are safety

related are provided in 10 CFR 50.2, "Definitions." Based upon these criteria, licensees'

engineering organizations develop plant-specific listings of safety-related SSCs.

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13.2.1 Appendix A to 10 CFR Part 50

Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 details the

general requirements for establishing QA controls. General Design Criterion 1, "Quality

Standards and Records," contains certain requirements applying to the QA of items important to

safety. The scope of items that are "important to safety" includes a subset of plant equipment

that is classified as safety related. Appendix B, "Quality Assurance Criteria for Nuclear Power

Plants and Fuel Reprocessing Plants," to 10 CFR Part 50 (discussed below) contains QA

program requirements for safety-related SSCs. Other regulatory guidance discusses

QA program controls that are appropriate for some types of nonsafety-related equipment.

13.2.2 Appendix B to 10 CFR Part 50

Appendix B to 10 CFR Part 50 outlines the QA requirements that apply to activities that affect

the safety-related functions of SSCs that prevent or mitigate the consequences of postulated

accidents. The appendix defines QA as all planned and systematic actions that are necessary

to provide adequate confidence that SSCs will perform satisfactorily in service. Toward that end,

Appendix B specifies 18 criteria that the commitments in a licensee's QA program must satisfy.

These criteria cover such aspects as organizational independence, design control, procurement,

document control, test control, corrective action, and audits. Appendix B also stipulates that

licensees establish measures to ensure that applicable regulatory requirements, design bases,

and other requirements that are necessary to ensure adequate quality are suitably included or

referenced in the documents for procurement of safety-related materials, equipment, and

services whether purchased by the licensee or its contractors or subcontractors. Consistent with

the importance and complexity of the products or services to be provided, licensees (or their

designees) are responsible for periodically verifying that suppliers' QA programs comply, as

appropriate, with the applicable criteria in Appendix B and that they are effectively implemented.

Additionally, as outlined in 10 CFR 21.41, the NRC staff performs inspections at vendors who

supply basic components to the nuclear industry.

The requirements of Appendix B are written at a conceptual level and it was necessary for the

NRC and the industry to develop consensus standards that include acceptable ways to conform

to these requirements. The NRC then issued companion regulatory guides, which endorsed

(with conditions, if warranted) QA codes and standards.

13.2.3 Approaches for Adopting More Widely Accepted International Quality Standards

As stated in the previous U.S. National Report, the NRC reviewed options for adopting more

widely accepted international quality standards like International Organization for

Standardization (ISO) Standard 9001, considering how international standards compare with the

existing Appendix B framework. On the basis of this review, the NRC concluded that

supplemental quality requirements would be needed when implementing ISO 9001 within the

existing regulatory framework. As part of the ongoing multinational design evaluation program,

the NRC will be reevaluating various international QA standards to achieve a greater degree of

international convergence.

13.3 QA Regulatory Guidance

The NRC has QA guidance for design and construction, operational activities, and licensing.

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13.3.1 Guidance for Design and Construction Activities

As discussed in the previous U.S. National Report, licensees apply consensus standards

developed by the American National Standards Institute (ANSI) in its N45.2 series or ASME in its

NQA-1 series as acceptable ways of complying with the requirements of Appendix B to

10 CFR Part 50. The NRC endorses ANSI and ASME standards through its regulatory guides.

The NRC staff, as part of consensus codes and standards activities, continues to participate with

ASME NQA-1 committees to revise the latest edition of the NQA-1 standard to be endorsable

within the 10 CFR 50.55(a) regulatory framework. Currently, a number of licensees have

implemented QA programs applying ASME NQA-1 -1994.

13.3.2 Guidance for Operational Activities

As described in the previous U.S. National Report, the NRC has conditionally endorsed the

consensus standard ANSI N18.7-1976, "Administrative Controls and Quality Assurance for the

Operational Phase of Nuclear Power Plants" through Regulatory Guide 1.33, "Quality Assurance

Program Requirements (Operations)," Revision 3, issued November 1980, as complying with the

requirements of Appendix B.

13.3.3 Guidance for Staff Reviews for Licensing

NUREG-0800 guides the staff review of applications. NUREG-0800 has specific review

guidance correlated with the 18 criteria of Appendix B to 10 CFR Part 50 and integrates a review

of licensee commitments to adopt the NRC's QA-related regulatory guides and the industry's QA

codes and standards. The draft of-NUREG-0800, Section 17.5, provides the NRC's

expectations for a QA program.

13.4 QA Progirams

The NRC inspects QA programs under the Reactor Oversight Process for operating reactors

and under the Construction Inspection Program for new reactors. Additionally, the NRC

conducts augmented inspection programs as needed.

The baseline inspection program of the Reactor Oversight Process includes one primary

procedure related to QA issues, known as Inspection Procedure 71152, "Identification and

Resolution of Problems." Inspectors use this procedure to assess the effectiveness of licensees'

programs to identify and resolve problems according to a performance-based review of specific

issues. In particular, inspectors look for cases in which a licensee may have missed generic

implications of specific problems and for the risk significance of combinations of problems that

individually may not have significance. They also verify that licensees are properly capturing

issues that could affect the availability of equipment that is tracked under 10 CFR 50.65,

"Maintenance Rule," or by performance indicators. They do not inspect other aspects of QA

program implementation in thebaseline inspection program, but may, through supplemental

inspections.

Some equipment in the nuclear facility may be classified as nonsafety related and yet still be

important to safety for some unique reason. In specific cases, the NRC has specified that QA

controls are warranted for equipment determined to be more important than commercial-grade

equipment. However, the QA controls do not have to meet Appendix B requirements, which

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apply only to activities affecting safety-related functions. Typically, applying QA controls to this

important-to-safety, yet nonsafety-related, equipment is called "augmented quality control."

The NRC's inspection program has been anticipating applications for combined licenses to

facilitate the construction of new nuclear power plants under 10 CFR Part 52. (See Article 18.)

The Construction Inspection Program provides oversight for future nuclear plants licensed under

10 CFR Part 52, including QA program inspection. The QA inspection program focuses on an

applicant establishing and implementing a QA program in accordance with Appendix B to 10

CFR Part 50 to ensure compliance with specified requirements in 10 CFR Part 52.

As provided in the Construction Inspection Program (see Article 18), the plant will transition from

the Construction Inspection Program to the Reactor Oversight Process for commercial operation

after the completion of startup open items.

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ARTICLE 14. ASSESSMENT AND VERIFICATION OF SAFETY

Each Contracting Party shall take the appropriate steps to ensure that:

(I) comprehensive and systematic safety assessments are carried out before the

construction and commissioning of a nuclear installation and throughout its life.

Such assessments shall be well documented, subsequently updated in the light of

operating experience and significant new safety information, and reviewed under

the authority of the regulatory body

(ii) verification by analysis, surveillance, testing, and inspection is carried out to

ensure that the physical state and the operation of nuclear installations continue

to be in assurance with its design, applicable national safety requirements, and

operational limits and conditions

This section explains the governing documents and process for ensuring that systematic safety

assessments are carried out during the life of the nuclear installation, including for the period of

extended operation. It focuses on assessments performed to maintain the licensing basis of a

nuclear installation. Finally, this section explains verification of the physical state and operation

of the nuclear installation by analysis, surveillance, testing, and inspection.

Other articles (e.g., Articles 6, 10, 13, 18, and 19) also discuss activities undertaken to achieve

safety at nuclear installations.

Updates to this section include information on the amendment process, Browns Ferry restart,

and license conditions.

14.1 Ensuring Safety Assessments throughout Plant Life

Before a nuclear facility is constructed, commissioned, and licensed, an applicant must perform

comprehensive and systematic safety assessments, which are reviewed and approved by the

NRC. Article 18 discusses these assessments and reviews. This section focuses on the

assessments that are required throughout the life of a nuclear installation (i.e., assessments

required to maintain the licensing basis). To show conformance with the licensing basis, a

licensee must maintain records of the original design bases and any changes. This section

explains how such changes are documented, updated, and reviewed. Renewal of a license is

predicated on the requirement that a licensee will continue to meet its current licensing basis;

this section explains how the license renewal process accounts for this requirement.

14.1.1 Maintaining the Licensing Basis

The NRC carries out regulatory programs to give reasonable assurance that plants continue to

conform to the licensing basis. (Article 6 discusses these programs.) This section explains the

governing documents and process used to maintain the licensing basis. The main governing

documents are 10 CFR 50.90, "Application for Amendment of License or Construction Permit,"

10 CFR 50.59, "Changes, Tests, and Experiments," and 10 CFR 50.71, "Requirements for

Updating of Final Safety Analysis Reports."

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14. 1.1.1 Governing Documents and Process

A licensee is to operate its facility in accordance with the license and as described in its final

safety analysis report. To change its license or reactor facility, a licensee must follow the review

and approval processes established in the regulations. For license amendments, including

changes to technical specifications, the licensee must request NRC approval in accordance with

10 CFR 50.90. However, 10 CFR 50.59 (as described below) contains requirements for the

process by which licensees may make changes to their facilities and procedures as described in

the safety analysis report, without prior NRC approval, under certain conditions.

10 CFR 50.59. 10 CFR 50.59 establishes the conditions under which licensees may make

changes to the facility or procedures and conduct tests or experiments without prior NRC

approval. Proposed changes, tests, and experiments that satisfy the definitions and one or more

of the criteria in the rule must be reviewed and approved by the NRC before implementation.

Thus the rule provides a threshold for regulatory review, not the final determination of safety, for

proposed activities. After determining that a proposed activity is safe and effective through

appropriate engineering and technical evaluations, the 10 CFR 50.59 process is applied to

determine if a license amendment is required prior to implementation. The process involves

three basic steps: (1) applicability and screening to determine if a 10 CFR 50.59 evaluation is

required; (2) an evaluation that applies the eight evaluation criteria of 10 CFR 50.59 (c)(2) to

determine if a license amendment must be obtained from the NRC; and, (3) documentation and

reporting to the NRC of activities implemented under 10 CFR 50.59.

A licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a

proposed change, test, or experiment if the change, test, or experiment would: result in more

than a minimal increase in the frequency of occurrence of a previously evaluated accident;

result in more than a minimal increase in the likelihood of occurrence of a malfunction of a

I structure, system, or component (SSC) important to safety; result in more than a minimal

increase in the consequences of a previously evaluated accident; result in more than a minimal

increase in the consequences of a malfunction of an SSC important to safety; create a possibility

for an accident of a different type than any previously evaluated; create a possibility for a

malfunction of an SSC important to safety with a different result than any previously evaluated;

result in a design basis limit for a fission product barrier being exceeded or altered; or result in a

departure from a method of evaluation used in establishing the design bases or in the safety

analyses. According to § 50.90, whenever a holder of a license or construction permit desires to

amend the license or permit, it must file an application for an amendment with the Commission,

as specified in § 50.4, fully describing the changes desired, and following as far as applicable,

the form prescribed for original applications. The NRC performs and documents a safety

evaluation in these instances before it authorizes the change.

10 CFR 50.71. Paragraph (e) of 10 CFR 50.71 sets forth another process for making changes.

This regulation requires licensees to update their final safety analysis reports periodically to

incorporate the information and analyses that they submitted to-the Commission or prepared

pursuant to Commission requirements. Revisions to the updated final safety analysis reports

are to include the effects of changes that occur in the vicinity of the plant, changes made in the

facility or procedures described in the report, safety evaluations for approved license

amendments and for changes made under 10 CFR 50.59, and safety analyses conducted at the

request of the Commission to address new safety issues.

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14.1.1.2 Regulatory Framework for the Restart of Browns Ferry Unit 1

This section describes the assessment and verification of safety for a plant that was restarted

after being shutdown for a number of years.

As background, the Browns Ferry site, located near Decatur, Alabama, has three BWRs

(General Electric, BWR-4, Mark-1 containment). All three units were shut down in 1985 to

address management and regulatory issues. After resolving these issues, TVA successfully

restarted Units 2 and 3 in the 1990s' but kept Unit 1 in a defueled layup condition. In May 2002,

TVA decided to initiate a restart effort for Unit 1. The three Browns Ferry units are similar in

design and licensing basis. TVA has implemented programs for Unit 1 that are similar to those

used to restart Units 2 and 3, incorporated improvements and lessons learned and dedicated

resources, including personnel with experience on restarting Units 2 and 3. The restart of Unit 1

differed from the restart of Units 2 and 3 in that TVA applied simultaneously for both a license

renewal and an extended power uprate for Unit 1.

The regulatory framework for the restart of Browns Ferry Unit 1 consisted of two major

elements-inspection and licensing activities. The NRC performed inspections in accordance with

Inspection Manual Chapter 2509 and conducted the licensing activities consistent with the

NRC's August 2003 regulatory framework letter discussed below.

TVA has submitted many varied licensing actions. The regulatory framework letter, provided a

detailed listing of generic communications and other licensing actions requiring regulatory review

and approval as wells as follow-up inspection. To facilitate communications with key

stakeholders, the NRC held periodic public meetings at the site and developed a public outreach

Web page which is similar to the Reactor Oversight Process Web page.

As part of its inspection program, the NRC staff reviewed TVA programs and plant activities

related to the recovery of Unit 1. These activities included replacement, renovation, and removal

of equipment and a review of plant programs, process and training of plant personnel. The NRC

inspections of structural, electrical, mechanical and fire protection modifications resulted in

satisfactory findings. Onsite monitoring and review determined that activities involving

replacement, renovation, and removal of equipment were satisfactorily conducted so as to

maintain adequate nuclear and radiological safety.

In addition, the NRC conducted an Operational Readiness Assessment Team inspection in April

2007 to assess management controls, implementation of site programs and personnel readiness

to support safe restart and operation of Unit 1. The purpose of this inspection was to focus on

the effectiveness of licensee management oversight, safety-significant activities, operator

training and experience, corrective action programs, the maintenance program, operator

response to annunciators and general plant conditions impacting safety, and the readiness to

support three-unit operations. The inspection concluded that site programs, personnel and

procedures were adequate for restart of Unit 1 and three unit power operations.

On May 15, 2007, the .NRC authorized TVA to restart Browns Ferry Unit 1. The unit was

restarted on May 22, 2007 and reached 100 percent power on June 8, 2007. TVA will conduct

post restart testing, including the performance of two large transients tests, scheduled for

completion within a few months.

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After extensive reviews and inspections over the past several years, and completion of the

regulatory framework issues, the NRC oversight for all cornerstones is now being conducted in

accordance with the reactor oversight process. However, due to the lack of valid historical plant

specific data for the remaining three cornerstones, additional reactor oversight process baseline

inspections will be necessary until sufficient plant specific data becomes available.

14.1.2 License Renewal

This section explains license renewal. It discusses the governing documents and regulatory

process, and recent experience, and provides relevant examples.

14.1.2.1 Governing Documents and Process

Background. The Atomic Energy Act and NRC regulations limit commercial power reactor

licenses to 40 years but permit such licenses to be renewed. The original 40-year term was

selected on the basis of economic and antitrust considerations, not technical limitations.

The NRC has established a license renewal process that can be completed in a reasonable

period of time with clear requirements to ensure safe plant operation for up to an additional

20 years of plant life. The NRC's current schedule is to complete renewal reviews within

30 months of receipt of the application if a hearing is conducted, and within 22 months if not.

The NRC completes nonstandard applications according to a schedule agreed upon with the

applicant. The decision to seek license renewal rests entirely with nuclear power plant owners

and typically is based on the plant's economic situation and whether it can meet NRC

requirements.

Research has concluded that aging phenomena are readily manageable and do not pose

technical issues that would preclude life extension for nuclear power plants. Studies have also

found that facilities deal adequately with many aging effects during the initial license period, and

credit should be given for these existing programs, particularly those under NRC's Maintenance

Rule (10 CFR 50.65), which helps manage plant aging.

The license renewal process proceeds along two tracks-one for review of safety issues and

another for environmental issues. An applicant must provide the NRC with an evaluation that

addresses the technical aspects of plant aging and describes the ways it will manage those

effects. It must also prepare an evaluation of the potential impact on the environment if the plant

operates for up to another 20 years. The NRC reviews the application and verifies the safety

evaluations through inspections.

Public participation is an important part of the license renewal process. Members of the public

have opportunities to question how aging will be managed. during the period of extended

operation, and all information related to the review and approval of a renewal application is

publicly available. Significant concerns may also be litigated in an adjudicatory hearing if any

party who would be adversely affected requests a hearing.

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10 CFR Part 54, "Requirements for Renewal of Operating Licenses for Nuclear Power Plants."

Known as the License Renewal Rule, 10 CFR Part 54 establishes the technical and procedural

requirements for renewing operating licenses. License renewal requirements for power reactors

are based on two key principles:

(1) When continued into the extended period of operation, the regulatory process, which

assesses and verifies safety, is adequate to ensure that the licensing basis of all

currently operating plants provides an acceptable level of safety. The possible exception

is detrimental effects of aging on certain SSCs, and possibly a few other issues applying

to safety only during the period of extended operation.

(2) Each plant must maintain its licensing basis throughout the renewal term.

Guidance that applies to license renewal includes Regulatory Guide 1.188, "Standard Format

and Content for Applications to Renew Nuclear Power Plant Operating Licenses," issued July

2001, to assist applicants in applying to renew a license and NUREG-1800, which guides the

staff in reviewing applications. The Standard Review Plan incorporates by reference NUREG1801, "Generic Aging Lessons Learned (GALL) Report," issued July 2001, which generically

documents the basis for determining when existing programs are adequate and when they

should be augmented for license renewal. As lessons are learned from the review of renewal

applications or generic technical issues are resolved, the NRC issues improved guidance for

interim use by applicants until the guidance is incorporated into the next formal update of the

documents.

10 CFR Part 51, "Environmental Protection Regulations for Domestic Licensing and Related

Regulatory Functions." The NRC's environmental protection regulation, 10 CFR Part 51, also

applies to license renewal. This regulation was amended to facilitate the agency's

environmental review process for license renewal. Specifically, the review requirements for

10'CFR Part 51 are founded on the conclusion that certain environmental issues can be

resolved generically and need not be evaluated in each plant-specific application. These issues

are described in NUREG-1437, "Generic Environmental Impact Statement for License Renewal

of Nuclear Plants," issued May 1996. The NRC performs plant-specific reviews of the

environmental impacts of license renewal to determine whether the impacts are so great that

they should preclude license renewal as an option for energy-planning decisionmakers.

Supplement 1 to NRC Regulatory Guide 4.2, "Preparation of Supplemental Environmental

Reports for Applications to Renew Nuclear Power Plant Operating Licenses," issued August

1991, provides guidance to applicants preparing environmental reports for license renewal.

NUREG-1 555, "Standard Review Plans for Environmental Reviews for Nuclear Power Plants,

Supplement 1: Operating License Renewal," issued March 2000, guides the NRC staff's review

of the environmental issues associated with a renewal application

14.1.2.2 Experience

The NRC issued the first renewed licenses for the Calvert Cliffs Nuclear Power Plant and the

Oconee Nuclear Station in 2000. On the basis of industry statements, the NRC expects that

essentially all remaining plants will apply for license renewal.

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14.1.3 The United States and Periodic Safety Reviews

This report has retained the section on periodic safety reviews (PSRs) because of widespread

interest in this topic at the Convention's third review meeting.

This section explains how the U.S. regulatory approach provides a continuum of assessment

and review that ensures public health and safety throughout the period of plant operation. As

discussed below, plant safety is improved by a combination of the ongoing NRC regulatory

-process, oversight of the current licensing basis, backfitting, broad-based evaluations, license

renewal, and licensee initiatives that go beyond the regulations.

14.1.3.1 The NRC's Robust and Ongoing Regulatory Process and the Current Licensing Basis

Before issuing an operating license, the NRC comprehensively determines that the design,

construction, and proposed operation of the nuclear power plant satisfy the NRC's requirements

and reasonably ensure the adequate protection of the public health and safety. However,

the licensing basis of a plant does not remain fixed for the 40-year term of the operating license.

The licensing basis evolves throughout the term of the operating license because of the

continuing regulatory activities of the NRC, as well as the activities of the licensee.

The NRC engages in many regulatory activities which, when considered together, constitute a

process that provides ongoing assurance that the licensing bases of nuclear power plants

provide an acceptable level of safety. This process includes inspections (both periodic regional

inspections as well as daily oversight by the resident inspector), audits, investigations,

evaluations of operating experience, regulatory research, and regulatory actions to resolve

identified issues. The NRC's activities may result in changes to the licensing basis for nuclear

power plants through promulgation of new or revised regulations, acceptance of licensee

commitments to modify nuclear power plant designs and procedures, and the issuance of orders

or confirmatory action letters. The agency also publishes the results of operating experience

analysis, research, or other appropriate analyses through generic communication documents

such as bulletins and generic letters. Licensee commitments in response to these documents

also change the plant's licensing basis. In this way, the NRC's consideration of new information

provides ongoing assurance that the licensing basis for the design and operation of all nuclear

power plants provides an acceptable level of safety. This process continues for plants that

receive a renewed license to operate for 20 years beyond the original operating license.

In addition to NRC-required changes in the licensing basis, a licensee may also voluntarily seek

changes to the current licensing basis for its plant. However, these changes are subject to the

NRC's formal regulatory controls on changes (such as 10 CFR 50.54; 10 CFR 50.59;

10 CFR 50.90; and 10 CFR 50.92, "Issuance of Amendment"). These regulatory controls

ensure that licensee-initiated changes to the licensing basis for a plant are documented and that

the licensee obtains NRC review and approval before implementing changes to the licensing

basis that meet the review thresholds in 10 CFR 50.59. The licensee must report to the NRC

any changes or modifications it makes to the licensing basis without prior NRC review at least

every 2 years. Region-based NRC inspectors perform a sampling inspection of those changes

in accordance with the Reactor Oversight Process to ensure that the licensee has properly

characterized the changes or modifications.

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14.1.3.2 The Backfitting Process: Timely Imposition of New Requirements

The NRC recognized the need to consider new requirements systematically rather than

depending on the license renewal process or other regulatory processes to decide on plant

upgrades. In the late 1970s and early 1980s, the NRC recognized the need for a process to

determine when to address generic issues for all plants. As a result, the NRC developed the

"backfitting" process and initiated the Committee To Review Generic Requirements (CRGR) to

review staff-proposed backfits on licensees.

Also known as the Backfit Rule, 10 CFR 50.109 applies to both generic and plant-specific

backfits for power reactors. It defines a "backfit" as any modification of or addition to plant

systems, structures, components, procedures, organizations, design approvals, or

manufacturing licenses that may result from the imposition of a new or amended rule or

regulatory staff position. Except in the case of backfits that are imposed to bring a licensee back

into compliance with its license or to ensure adequate protection of the public health and safety

or common defense and security, the rule requires a cost-benefit backfit analysis. The NRC

must determine though a backfit analysis that the proposed backfit will substantially increase the

overall protection of the public health and safety (or common defense and security) and that the

direct and indirect costs for the facility are justified in view of the increased protection.

Compliance and adequate protection backfits are justified differently. A documented evaluation

is required, which provides the basis and states the objectives and purpose of the proposed

backfit.

In 1988, the NRC issued an amended Backfit Rule, which clearly states that economic costs will

not be considered in cases of ensuring, defining, or redefining adequate protection of the public

health and safety, or in cases of ensuring compliance with NRC requirements or written licensee

commitments.

Backfitting is expected to occur and is an inherent part of the regulatory process. However, it is

permitted only after a formal, systematic review to ensure that changes are properly justified and

suitably defined. The requirements of this process are intended to ensure order, discipline, and

predictability and to optimize use of NRC staff and licensee resources.

The controls on generic backfitting include review by the CRGR, a committee of senior

managers from various NRC offices. Established in 1981, this committee operates under a

charter that specifically identifies the documents to be reviewed and the analyses, justifications,

and findings to be provided. Its objectives include eliminating unnecessary burdens on

licensees, reducing radiation exposure to workers while implementing requirements, and

optimizing use of NRC and licensee resources to ensure safe operation. Thus, the CRGR

charter is a key implementing procedure for generic backfitting, although the primary

responsibility for proper backfit considerations belongs to the initiating organization.

14.1.3.3 The NRC's Extensive Experience with Broad-Based Evaluations

In the mid-1970s, the NRC recognized the importance of assessing the adequacy of the design

and operation of currently licensed nuclear power plants, understanding the safety significance

of deviations from applicable current safety standards that may have been approved after those

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plants were licensed, and providing the capability to make integrated and balanced decisions

about the need for backfit modifications at those plants.

Consequently, in 1977, the NRC initiated the Systematic Evaluation Program (SEP). From a list

of approximately 800 potential issues and topics related to nuclear safety, the SEP found that

the regulatory requirements for 137 issues had changed sufficiently to warrant evaluation. The

staff then compared the designs of 10 of the older plants to the licensing criteria delineated in

the then recently issued Standard Review Plan.3 After further review, the staff determined that

27 issues required some corrective action at one or more plants and resolution of those issues

could lead to safety improvements at other operating plants built at about the same time. These

27 issues became known as the 27 "SEP lessons learned."

In 1984, NRC staff presented the 27 SEP lessons learned to the Commission as part of a

proposal for an Integrated Safety Assessment Program (ISAP). The staff developed this

program to review safety issues for a specific plant in an integrated manner instead of continuing

the SEP at other older operating reactors. In November 1984, the Commission published the

"Commission Policy Statement on the Systematic Evaluation of Operating Nuclear Power

Reactors." In this policy statement, the Commission articulated its view that issues relating to

the safety of operating nuclear power plants can be more effectively and efficiently implemented

in an integrated, plant-specific review. For the first time, the Commission discussed probabilistic

safety analysis as a method to obtain consistent and comparable results which could be used to

enhance a safety assessment. The SEP process was transformed into the ISAP pilot program.

In May 1985, the NRC initiated the ISAP pilot program at two plants, Millstone Unit 1 and

Haddam Neck (Connecticut Yankee). The ISAP pilot program identified some benefits;

however, the Commission deferred extending it beyond the pilot phase until the staff provided an

integrated package of options that clarified the relationship between the proposed follow-on

program to the ISAP pilot (ISAP II) and the newly proposed individual plant examination (IPE)

process.

The Commission determined that, since the ISAP II program would be voluntary and the IPE

program, through the NRC's generic letter process, would require a licensee response, it should

give priority to the IPE program. Many of the same benefits that might have been derived

through the proposed ISAP II were derived instead through the IPE (e.g., probabilistic safety

analysis) process.

In the late 1980s and throughout the 1990s, the NRC continued its efforts to strengthen its

regulatory infrastructure and ensure continued safe operation of commercial nuclear power

plants through inspection, broad-based assessment, and where appropriate, establishment of

new generic requirements. For example, the Commission determined that licensees should

assess the accessibility and adequacy of their design-basis information and determine whether

their plants needed a design-basis reconstitution program. The Commission expressed its

expectations in "Availability and Adequacy of Design Bases Information at Nuclear Power Plants;

Standard Review Plans help ensure the quality and uniformity of staff reviews and provide a well-defined

base from which to evaluate a licensee or applicant submittal. The Standard Review Plans are also

intended to make information about regulatory matters widely available, to enhance communication with

interested members of the public and the nuclear power industry, and to improve the understanding of the

staff review process.

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Policy Statement," in the Federal Register of August 10, 1992. The Commission also expanded

the IPE program to consider external events and, recognizing the relationship between

maintenance, equipment reliability, plant risk, and safety, the Commission promulgated the

Maintenance Rule.

14.1.3.4 License Renewal Confirms Safety of Plants

As late as 1991, some plants had not definitively resolved the 27 SEP lessons learned. As the

staff considered a process to renew the operating licenses for the operating nuclear power

plants, it assessed the best way to address these 27 issues.

Of the 27 issues, four had been completely resolved for all plants. One other issue was of such

low safety significance that it required no additional action. The staff determined that none of

the remaining 22 issues required immediate action to protect public health and safety. The staff

placed these 22 issues into the established regulatory process for determining the safety

significance of generic issues.4

In developing the License Renewal Rule, the Commission concluded that issues material to the

renewal of a nuclear power plant operating license are limited to those issues that the

Commission determines are uniquely relevant to protecting the public health and safety and

preserving common defense and security during the period of extended operation. Other issues

would, by definition, be relevant to the safety and security of the public during current plant

operation. Given the Commission's ongoing obligation to oversee the safety and security of

operating reactors, the existing regulatory process within the present 40-year license term would

address issues related to current plant operation rather than deferring the issues until the time of

license renewal. (See Section 14.1.2 for a description of license renewal.) To add to that

discussion, license renewal applicants are required to complete an integrated plant assessment

(IPA) 5 and evaluate time-limited aging analyses.

14.1.3.5 Risk-Informed Regulation and the Reactor Oversight Process

The NRC is actively increasing the use of risk insights and information in its regulatory decisionmaking. In the reactor area, risk-informed activities occur in the five broad categories of (1)

applicable regulations, (2) licensing process, (3) revised oversight process, (4) regulatory

A generic issue is a regulatory matter that is not sufficiently addressed by existing regulations, guidance, or programs. The NRC has identified by its systematic assessment of plant operation certain issues that

seem prevalent among plants. The NRC documents and tracks resolution of these "generic safety issues."

The generic safety issue program provides for (1) identifying generic issues, (2) assigning them priorities,

(3) developing detailed action plans for their resolution, (4) overseeing progress in their resolution by senior

managers, and (5) informing the public of the status of progress in resolution. The resolution of these

issues may involve new or revised rules, new or revised guidance, or revised interpretation of rules or

guidance that affect nuclear power plant licensees or nuclear material certificate holders. Congress

requires that the NRC maintain this program.

An IPA identifies and lists structures and components subject to an aging management review. These

include "passive" structures and components that perform their intended function without moving parts or without a change in configuration or properties. Examples of these are components such as the reactor

vessel, the steam generators, piping, component supports, and seismic Category I structures. To be in

scope, the item must also be "long-lived" to be considered during the license renewal process. Long-lived means the item is not subject to replacement based on a qualified life or specified time period.

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guidance, and (5) risk analysis tools, methods, and data. Activities within these categories

include revisions to technical requirements in the regulations; risk-informed technical

specifications; a new framework for inspection, assessment, and enforcement actions; guidance

'on risk-informed inservice inspections; and improved standardized plant analysis risk models.

In 2000, the NRC implemented a revised Reactor Oversight Process (see discussion under

Article 6) using risk insights and lessons learned from more than 40 years of regulating nuclear

power plants. The previous oversight process evolved during a period when the nuclear power

industry was less mature and there was much less operational experience on which to base

rules and regulations. Very conservative judgments governed the rules and regulations.

Significant plant operating events occurred with some frequency, and the oversight process

tended to be reactive and prescriptive, closely observing plant performance for adherence to the

regulations and responding to operational problems as they occurred.

After nearly four decades of operational experience and generally steady improvements in plant

performance, the Reactor Oversight Process now focuses more of the agency's resources on

the relatively small number of plants that evidence performance problems. The Reactor

Oversight Process is more effective in correcting performance or equipment problems today

because the agency's response to problems is more timely and predictable.

The Reactor Oversight Process makes greater use of objective performance indicators.

Together, the indicators and inspection findings provide the information needed to support

reviews of plant performance, which are conducted quarterly. In addition, the Reactor Oversight

Process features expanded semiannual reviews, which include inspection planning and a

performance report (all of which are posted on the NRC's public Web site).

14.1.3.6 Licensee Responsibilities for Safety: Regulations and Initiatives Beyond Regulations

As in many countries, U.S. nuclear power plant licensees are responsible for the safety of their

facilities. This responsibility is embedded in their license and in the NRC's regulatory

infrastructure. Under the regulatory umbrella, licensees routinely assess new technologies, offnormal conditions, operating experience, and industry trends to make informed decisions about

safety enhancements to their facilities.

The NRC does not specifically mandate some of these reviews. Rather, they are self-imposed

initiatives over and above regulations, motivated by the licensees' self-described pursuit of

excellence and by the recognition that, in the U.S. free-market competitive energy industry,

safety and economics are directly linked. Licensees have, for example, voluntarily replaced

analog instrumentation and control systems with digital systems, upgraded their plants to

increase production of electricity, and managed their plants to performance levels above the

NRC's performance indicator thresholds.

Under the U.S. regulatory structure, Appendix B to 10 CFR Part 50 requires that all nuclear

power plant licensees maintain a QA program. QA comprises all those planned and systematic

actions necessary to provide adequate confidence that an SSC will perform satisfactorily in

service. QA includes quality control, which comprises those QA actions related to the physical

characteristics of a material, structure, component, or system that provide a means to control

quality to predetermined requirements.

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Licensees carry out a comprehensive system of planned and periodic audits to verify compliance

with all aspects of the QA program and to determine the effectiveness of the program.

Appropriately trained personnel who do not have direct responsibilities in the areas being

audited perform these audits in accordance with written procedures or checklists. Audit results

are documented and reviewed by management with responsibility in the area audited, and

appropriate followup is initiated.

14.1.3.7 Summary

IAEA and the Western European Nuclear Regulators' Association (WENRA) have developed

guidance6 and objectives for conducting PSRs that have much in common. Specifically,

consistent with the guidance of both organizations, PSRs are comprehensive assessments with

the following purposes:

to determine, at the time of the review, whether the plant complies with its licensing basis

to identify the extent to which the current licensing basis remains valid, in part by

determining the extent to which the plant meets current safety standards and practices

to provide a basis for implementing appropriate safety improvements, corrective actions,

or process improvements

to provide confidence that the plant can continue to be operated safely

For the reasons discussed above and summarized below, the shared objectives associated with

IAEA and WENRA PSR guidance are substantively accomplished in the United States on an

ongoing basis.

First, the NRC's regulatory process provides a robust foundation for ongoing assessments,

evaluations, and when appropriate, imposition of new requirements. The NRC and the U.S.

nuclear industry consider new information, in a more risk-informed manner, as it becomes

available; adjust the regulatory oversight and plant safety priority, respectively; and provide

ongoing assurance that the licensing basis for the design and operation of all nuclear power

plants provides an acceptable level of safety,

Second, the NRC and the U.S. nuclear industry have a 30-year history of implementing broadbased plant assessments. The regulatory history of implementing broad-based assessments is

a direct result of an adaptive, probing, and independent regulatory process. These assessments

have included the SEP, the ISAP, the IPE, and the reactor license renewal process and provide

additional confidence that plant safety continues to be the highest priority and that the NRC and

industry continue to pursue enhancements that improve safety. The time line in Figure 4

demonstrates that, over a period of almost 25 years, broad-based NRC assessments and

regulatory initiatives have provided a continuum of assessment, improvement, and oversight,.

which ensures that licensed plants continue to operate safely.

IAEA guidance appears in Safety Standards Series No. NS-G-2.10, "Periodic Safety Review of Nuclear

Power Plants Safety Guide,"issued in 2003. WENRA guidance appears in "Pilot Study on Harmonization of

Reactor Safety in WENRA Countries," WENRA Working Group on Reactor Harmonization, March 2003.

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The NRC's approach to continuing to ensure plant safety differs from the historically

deterministic focus of PSRs. The transition to a more risk-informed regulatory framework and

the Reactor Oversight Process provides an ongoing approach and basis for implementing

appropriate safety improvements, corrective actions, or process improvements and provides

confidence that the plant can continue to be operated safely. The NRC's more risk-informed

approach helps ensure that resources are optimally focused on those issues most important to

safety.

Finally, U.S. licensees establish performance expectations above the thresholds required by the

NRC. These self-imposed expectations and initiatives-over and above the regulations-result

from the licensee's self-described motivation to pursue excellence and by the recognition that, in

the free-market competitive industry in the United States, safety and economics are directly

linked.

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Timeline of Significant U.S. Regulatory Broad Based Reviews

1980 19901 20001

Integrated Safety

Assessment

Program

Indivi

Exam

IPE -

I~ 4i

Systematic Evaluation Process

| I

dual Plant Security and

ination (IPE) and Safeguards

External Events Comprehensive

Review

SIReconstitution Design Basis Revised Reactor

Oversight Process

Maintenance Rule 50.65 - risk-informed and performance based

Development of Backfitting Process

0

101

14.2 Verification bv Analysis. Surveillance. Testina. and Inspection

Licensees are required to verify that they are operating their nuclear installations in accordance

with the plant-specific design and requirements. The technical specifications (for surveillance)

and national consensus codes (for testing and periodic inspections) contain the requirements

specifying verification.

In 10 CFR 50.55a, "Codes and Standards," the requirements appear for applying industry codes

and standards to nuclear power reactors during design, construction, and operation. This

section states, "Systems and components of boiling and pressurized water-cooled nuclear power

reactors must meet the requirements of the ASME Boiler and Pressure Vessel Code specified in

paragraphs (b) through (g) of this section." In addition, 10 CFR 50.55a provides for alternatives

to the ASME Code when authorized by the NRC.

Through analysis, surveillance, testing, and inspection, the NRC verifies that the physical state

and operation of nuclear installations continue to be in accordance with the designs, applicable

national safety requirements, and operational limits and conditions. As previously discussed in

Article 6, the NRC's Reactor Oversight Process includes inspections to verify that licensees are

fulfilling their obligations to conduct such surveillances and testing and take corrective action.

The NRC staff updates, revises, and improves existing regulatory programs in light of operating

experience and significant new safety information. Article 19 discusses these activities.

The Commission may also require under 10 CFR 50.54(f) "Conditions of licenses" that licensees

submit written statements to enable the Commission to determine whether or not the license

should be modified, suspended, or revoked.

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