Final Safety Evaluation 

Report

Related to the Aircraft Impact 

Amendment to the U.S. Advanced 

Boiling Water Reactor (ABWR) 

Design Certification

Office of New Reactors

NUREG-1948
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Final Safety Evaluation 

Report

Related to the Aircraft Impact 

Amendment to the U.S. 

Advanced Boiling Water 

Reactor (ABWR) Design 

Certification

Manuscript Completed: October 2010

Date Published: June 2011

Office of New Reactors

NUREG-1948

iii

ABSTRACT 

This safety evaluation report (SER) documents the technical review of U.S. Advanced Boiling 

Water Reactor (ABWR) Aircraft Impact Assessment (AIA) application by the U.S. Nuclear 

Regulatory Commission (NRC) staff. The ABWR AIA application was submitted by the STP 

Nuclear Operating Company (STPNOC), in accordance with the procedures of Subpart B to 

Part 52 of Title 10 of the Code of Federal Regulations (10 CFR). 

The U.S. ABWR is a single-cycle, forced-circulation, boiling water reactor (BWR) with a rated 

power of 3926 megawatts thermal (MWt) and a design power of 4005 MWt. The staff issued the 

“Final Safety Evaluation Report (FSER) Related to Certification of the Advanced Boiling Water 

Reactor Design" (NUREG-1503), on July 13, 1994. The U.S. Nuclear Regulatory Commission 

issued a final rule certifying the design on May 12, 1997. 

On June 30, 2009, STPNOC submitted an application to amend the Design Certification Rule for 

the ABWR. On September 23, 2010, STPNOC submitted the final amendment application 

(ML102870017). The purpose of the amendment is to address the requirements of the NRC in 

10 CFR 50.150, “Aircraft Impact Assessment.” If the proposed amendment is approved, 

applicants for a combined license (COL) that reference the ABWR standard design may address 

the requirements of 10 CFR 50.150 by referencing the amended ABWR standard design. 

Under 10 CFR 50.150, applicants for new nuclear power reactors are required to perform an 

assessment of the effects on the designed facility of the impact of a large, commercial aircraft. 

Using realistic analyses, applicants must identify and incorporate into the design those 

design features and functional capabilities to show, with the reduced use of operator action, that 

(1) the reactor core remains cooled or the containment remains intact and (2) spent fuel cooling 

or spent fuel pool integrity is maintained (referred to as the acceptance criteria). Applicants 

subject to this regulation are required to submit a description of the design features and 

functional capabilities identified as a result of the AIA and a description of how those features 

and capabilities show that the acceptance criteria in 10 CFR 50.150(a)(1) are met with reduced 

use of operator action. 

The NRC determined that the impact of large, commercial aircraft is a beyond-design-basis 

event. Therefore, applicants may identify either safety-related or non-safety-related features or 

capabilities to satisfy the requirements for consideration of aircraft impact. The design features 

relied upon to satisfy the requirements for consideration of aircraft impact may be structures or 

features (1) whose sole purpose is to address these requirements, or (2) that have a dual 

purpose of addressing aircraft impact requirements as well as other NRC requirements. 

The NRC’s review of the applicant’s proposed amendment has three objectives. The first 

objective is to confirm that the applicant has (1) adequately described design features and 

functional capabilities in accordance with the aircraft impact rule; and (2) conducted an 

assessment reasonably formulated to identify design features and functional capabilities to show, 

with reduced use of operator action, that the facility can withstand the effects of an aircraft 

impact. This evaluation is documented in Chapter 19 of this safety evaluation report (SER). 
iv 

The second objective of the review is to determine that there will be no adverse impacts from 

complying with the requirements for consideration of aircraft impacts on conclusions reached by 

the NRC in its review of the original ABWR design certification (NUREG-1503). This evaluation 

is documented in Chapters 1, 3, 5, 6, 7, 9, 13, 14, 17, and 20 of this SER.` 

The third objective is to determine if the applicant is technically qualified in accordance with 

10 CFR 52.47(a)(7) to perform the design work to amend a portion of the ABWR design and to 

supply the amended portion of the design. This evaluation is documented in Chapter 1 of this 

SER. 

On the basis of its evaluation and independent analyses, the NRC staff concludes that 

STPNOC’s application meets the requirements of Subpart B of 10 CFR Part 52 and 

10 CFR 50.150(b). NRC staff issuance of this SER does not constitute a commitment to issue 

the design certification or in any way affect the authority of the Commission in any rulemaking 

proceeding pursuant to Subpart H of 10 CFR Part 2 and Subpart B to 10 CFR Part 52. 
v 

CONTENTS

ABSTRACT .................................................................................................................................. iii 

CONTENTS .................................................................................................................................. v

ABBREVIATIONS ....................................................................................................................... x 

1.0 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT ....................................... 1-1

1.1 Introduction ............................................................................................................ 1-1

1.2 Summary of Application ........................................................................................ 1-1

1.3 Regulatory Basis ................................................................................................... 1-2

1.4 Technical Evaluation ............................................................................................. 1-3

1.5 Conclusion ............................................................................................................. 1-5

3. DESIGN OF STRUCTURES, COMPONENTS, EQUIPMENT, AND SYSTEMS .............. 3-1

3.2.1 Seismic Classification ................................................................................ 3-1

 3.2.1.1 Introduction ............................................................................. 3-1

3.2.1.2 Summary of Application .......................................................... 3-1

3.2.1.3 Regulatory Basis .................................................................... 3-1

3.2.1.4 Technical Evaluation .............................................................. 3-1

3.2.1.5 Conclusion .............................................................................. 3-2

3.2.2 Quality Group Classifications .................................................................... 3-3

3.2.2.1 Introduction ............................................................................. 3-3

3.2.2.2 Summary of Application .......................................................... 3-3

3.2.2.3 Regulatory Basis .................................................................... 3-3

3.2.2.4 Technical Evaluation .............................................................. 3-3

3.2.2.5 Conclusion .............................................................................. 3-4

3.8.6 Alternate Feedwater Injection Pump House .............................................. 3-4

3.8.6.1 Introduction ............................................................................. 3-4

3.8.6.2 Summary of Application .......................................................... 3-4

3.8.6.3 Regulatory Basis .................................................................... 3-5

3.8.6.4 Technical Evaluation .............................................................. 3-5

3.8.6.5 Conclusion .............................................................................. 3-8

5 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS ................................... 5-1

5.2.2 Overpressure Protection ............................................................................ 5-1

5.2.2.1 Introduction ............................................................................. 5-1

5.2.2.2 Summary of Application .......................................................... 5-1

5.2.2.3 Regulatory Basis .................................................................... 5-1

5.2.2.4 Technical Evaluation .............................................................. 5-2

5.2.2.5 Conclusion .............................................................................. 5-2

6 ENGINEERED SAFETY FEATURES ............................................................................... 6-1

6.2.4 Containment Isolation System ................................................................... 6-1

6.2.4.1 Introduction ............................................................................. 6-1

6.2.4.2 Summary of Application .......................................................... 6-1
vi

6.2.4.3 Regulatory Basis .................................................................... 6-1

6.2.4.4 Technical Evaluation .............................................................. 6-1

6.2.4.5 Conclusion .............................................................................. 6-2

7 INSTRUMENTATION AND CONTROLS ......................................................................... 7-1

7.7 Instrumentation and Controls not Required for Safety ................................................ 7-1

7.7.1 Introduction ................................................................................................ 7-1

7.7.2 Summary of Application ............................................................................. 7-1

7.7.3 Regulatory Basis ....................................................................................... 7-2

7.7.4 Technical Evaluation ................................................................................. 7-2

7.7.5 Conclusion ................................................................................................. 7-3

9 AUXILIARY SYSTEMS..................................................................................................... 9-1

9.5.1 Fire Protection Program ............................................................................ 9-1

9.5.1.1 Introduction ............................................................................. 9-1

9.5.1.2 Summary of Application .......................................................... 9-1

9.5.1.3 Regulatory Basis .................................................................... 9-1

9.5.1.4 Technical Evaluation .............................................................. 9-1

9.5.1.5 Conclusion .............................................................................. 9-2

9.5.14 Alternate Feedwater Injection System ....................................................... 9-2

9.5.14.1 Introduction ............................................................................. 9-2

9.5.14.2 Summary of Application .......................................................... 9-2

9.5.14.3 Regulatory Basis .................................................................... 9-3

9.5.14.4 Technical Evaluation .............................................................. 9-3

9.5.14.5 Conclusion .............................................................................. 9-7

13 CONDUCT OF OPERATIONS ....................................................................................... 13-1

13.5.3.4 Procedures Included In Scope of Plant ................................ 13-1

13.5.3.4.1 Introduction ...................................................... 13-1

13.5.3.4.2 Summary of Application ................................... 13-1 

13.5.3.4.3 Regulatory Basis .............................................. 13-1

13.5.3.4.4 Technical Evaluation ........................................ 13-1

13.5.3.4.5 Conclusion ....................................................... 13-1

14 INITIAL TEST PROGRAM .............................................................................................. 14-1

14.2.1 Initial Test Program (ITP) ........................................................................ 14-1

14.2.1.1 Introduction .......................................................................... 14-1

14.2.1.2 Summary of Application ........................................................ 14-1

14.2.1.3 Regulatory Basis .................................................................. 14-2

14.2.1.4 Technical Evaluation ............................................................ 14-2

14.2.1.5 Conclusion ............................................................................ 14-2

14.3 Tier 1 Selection Criteria and Processes .............................................................. 14-3

14.3.1 Introduction .............................................................................................. 14-3

14.3.2 Summary of Application ........................................................................... 14-3

14.3.3 Regulatory Basis ..................................................................................... 14-3
vii

14.3.4 Technical Evaluation ............................................................................... 14-3

14.3.5 Conclusion ............................................................................................... 14-4 

17 QUALITY ASSURANCE................................................................................................. 17-1

17.1.19 Quality Assurance During the Design Phase ........................................ 17-1

17.1.19.1 Introduction ............................................................................. 17-1

17.1.19.2 Summary of Application ......................................................... 17-1

17.1.19.3 Regulatory Basis .................................................................... 17-1

17.1.19.4 Technical Evaluation .............................................................. 17-2

17.1.19.4.1 Organization ..................................................... 17-2

17.1.19.4.2 Quality Assurance Program .............................. 17-3

17.1.19.4.3 Design Control .................................................. 17-4

17.1.19.4.4 Procurement Document Control ....................... 17-4

17.1.19.4.5 Instructions, Procedures, and Drawings ........... 17-5 

17.1.19.4.6 Document Control ............................................. 17-5

17.1.19.4.7 Control of Purchased Material, Equipment, and 

Services ..................................................................... 17-6

17.1.19.4.8 Identification and Control of Materials, Parts, and 

Components ............................................................... 17-8

17.1.19.4.9 Control of Special Processes ........................... 17-8

17.1.19.4.10 Inspection ....................................................... 17-8

17.1.19.4.11 Test Control .................................................... 17-9

17.1.19.4.12 Control of Measuring and Test Equipment ..... 17-9

17.1.19.4.13 Handling, Storage, and Shipping .................... 17-9

17.1.19.4.14 Inspection, Test, and Operating Status .......... 17-9

17.1.19.4.15 Nonconforming Materials, Parts, or 

 Components ............................................................ 17-10

17.1.19.4.16 Corrective Action .......................................... 17-10

17.1.19.4.17 Quality Assurance Records .......................... 17-10

17.1.19.4.18 Quality Assurance Audits ............................. 17-11

17.1.19.4.19 Non-Safety-Related SSC Quality Assurance 

Controls .................................................................... 17-11

17.1.19.4.19.1 Non-Safety-Related SSCS – Significant 

Contributors to Plant Safety...............................17-11

17.1.19.4.19.2 Non-Safety-Related SSCS Credited for Regulatory 

Events...........................................................17-12

17.1.19.4.20 Regulatory Commitments ............................ 17-12

17.1.19.4.21 Additional Quality Assurance and Administrative 

Controls for the Plant Operational Phase ................. 17-13

17.1.19.5 Conclusion ............................................................................ 17-13 

19S DESIGN FEATURES FOR PROTECTION AGAINST A LARGE, COMMERCIAL 

 AIRCRAFT IMPACT .................................................................................................... 19S-1

19S.1 Introduction ....................................................................................................... 19S-1

19S.2 Summary of Application ................................................................................... 19S-1

19S.2.1 Description of Key Design Features .................................................. 19S-2

19S.2.2 Description of How Regulatory Acceptance Criteria are Met ............. 19S-2

19S.2.3 Inspection, Test, Analysis, and Acceptance Criteria (ITAAC) ............ 19S-3
viii

19S.3 Regulatory Basis .............................................................................................. 19S-4

19S.3.1 Review Guidance ............................................................................... 19S-4

19S.4 Technical Evaluation ........................................................................................ 19S-5

19S.4.1 Reasonably Formulated Assessment ................................................ 19S-6

19S.4.2 Key Design Features for Core Cooling .............................................. 19S-6

19S.4.3 Key Design Features that Protect Core Cooling Design Features ... 19S-11

19S.4.4 Key Design Features that Maintain Integrity of the Spent Fuel Pool 19S-14

19S.4.5 Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) ..... 19S-15

19S.5 Conclusion ...................................................................................................... 19S-17 

20 GENERIC ISSUES ......................................................................................................... 20-1

 20.5.1.3 Identification of Potential Design Improvements ............................. 20-1

20.5.1.3.1 Introduction ............................................................................ 20-1

20.5.1.3.2 Summary of Application ......................................................... 20-1

20.5.1.3.2 Regulatory Basis ................................................................... 20-1

20.5.1.3.4 Technical Evaluation ............................................................. 20-1

20.5.13.4 Conclusion .............................................................................. 20-2 
ix

APPENDICES 

Appendix A References .......................................................................................................... A-1 

Appendix B Chronology of Correspondence ....................................................................... B-1 

Appendix C List of Contributors ........................................................................................... C-1

Appendix D Chronology of Request for Additional Information ........................................ D-1 

Appendix E Report by Advisory Committee on Reactor Safeguards ................................ E-1 

Appendix F Staff Response to Advisory Committee on Reactor Safeguards Report ....... F-1
x 

ABBREVIATIONS 

A2LA American Association for Laboratory Accreditation

ABWR Advanced Boiling Water Reactor

ACLASS ACLASS Accreditation Services

ADS Automatic depressurization system 

AIA Aircraft impact assessment 

ASME American Society of Mechanical Engineers

ATWS Anticipated transient without SCRAM

COL Combined License 

COPS containment overpressure protection system 

CUW Reactor water cleanup system 

DC Design certification

DCD Design control document 

DG Draft regulatory guide 

ECCS Emergency core cooling systems 

ESP Early site permit

FSAR Final Safety Analysis Report

FSER Final Safety Evaluation Report 

GDC General Design Criterion/Criteria 

GPM Gallons per minute

HPCF High-pressure core flooder 

HPIN High pressure nitrogen

I&C Instrumentation and control 

IAS International Accreditation Service

IEEE Institute of Electrical and Electronic Engineers

IEEE Std IEEE Standard

ILAC International Laboratory Accreditation Cooperation

ITAAC Inspections, tests, analyses, and acceptance criteria

ITP Initial Test Program 

L-A-B Laboratory Accreditation Bureau

LPM Liters per minute

M&TE Measuring and test equipment

MOV Motor-operated valve 

MRA Mutual Recognition Arrangement

NEI Nuclear Energy Institute

NIRMA Nuclear Information and Records Management 

Association

NIST National Institute of Standards and Technology

NPSH Net positive suction head

NVLAP National Voluntary Laboratory Accreditation Program
xi 

P&ID Piping and instrumentation diagram

QA Quality Assurance 

QAPD QA Program Description

RAI Request for additional information 

RCIC Reactor core isolation cooling 

RCPB Reactor coolant pressure boundary 

RG Regulatory guide 

RPV Reactor pressure vessel

SBO Station blackout

SCG Startup Control Group

SER Safety evaluation report 

SRP Standard review plan 

SRV Safety/relief valve 

SSC Structures, systems and component 

SSE Safe-shutdown earthquake 

STPNOC

TS

STP Nuclear Operating Company

Technical specification 

TSD Technical Support Document

WR Wide range 

1-1

1.0 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 

1.1 Introduction 

As part of the application, STP Nuclear Operating Company (STPNOC) proposes changes to 

Chapter 1. Chapter 1, “Introduction and General Description of Plant,” of the Advanced Boiling 

Water Reactor (ABWR) design control document (DCD) describes the ABWR standard plant 

design, which includes all buildings primarily dedicated to housing the systems and the 

equipment related to the nuclear system, or access controls to the equipment and systems. 

The ABWR design comprises an essentially complete nuclear power plant except for sitespecific elements. This section of the safety evaluation report (SER) discusses the staff’s 

evaluation of the proposed revisions to Chapter 1 and the technical bases for the staff’s 

acceptance of these changes. 

1.2 Summary of Application 

On June 30, 2009, STPNOC (the applicant) submitted an application to amend the Design 

Certification Rule for the U.S ABWR. That rule approved Revision 4 of the U.S. ABWR DCD. 

Based on the staff’s Requests for Additional Information (RAIs) and the applicant’s internal 

reviews, STPNOC revised this application on May 12, 2010 (U7-C-STP-NRC-1000098, 

ML101340548), June 17, 2010, (U7-C-STP-NRC-100139, ML101720306), July 12, 2010 

(U7-C-STP-NRC-100168, ML102000496), August 4, 2010 (U7-C-STP-NRC-100187, 

ML102240435), and September 2, 2010 (U7-C-STP-NRC-100202, ML103190120). On 

September 23, 2010, STPNOC submitted the final application (U7-C-STP-NRC-100213, 

ML102870017). In these submittals, the applicant proposes the following changes to the 

certified DCD, Tier 2: 

1. Revised Section 1.1.1, “Format and Content,” to add a statement to address the location 

of the information added to comply with Title 10 of the Code of Federal Regulations

(10 CFR) 50.150. 

2. Revised Section 1.1.4, “Design Process,” to add a statement to address the design process 

for STPNOC amendment. 

3. Added Subsection 1.2.2.12.23, “Alternate Feedwater Injection (AFI) System,” to reference 

new Section 9.5.14, where the system is summarized. 

4. Added Subsection 1.2.2.16.16, “Alternate Feedwater Injection (AFI) Pump House,” to 

describe the location of the AFI Pump House and the equipment housed in the AFI pump 

house. 

5. Revised Figure 1.2-4, “Reactor Building, Arrangement Plan at Elevation 8200 mm” to reflect 

revisions proposed by the applicant. 

6. Revised Figure 1.2-5, “Reactor Building, Arrangement Plan at Elevation 1700 mm” to reflect 

revisions proposed by the applicant. 

7. Revised Figure 1.2-6, “Reactor Building, Arrangement Plan at Elevation 4800/8500 mm” to 

reflect revisions proposed by the applicant. 
1-2

8. Revised Figure 1.2-8, “Reactor Building, Arrangement Plan at Elevation 12300 mm” to 

reflect revisions proposed by the applicant. 

9. Revised Figure 1.2-9, “Reactor Building, Arrangement Plan at Elevation 18100 mm” to 

reflect revisions proposed by the applicant. 

10. Revised Figure 1.2-10, “Reactor Building, Arrangement Plan at Elevation 23500 mm” to 

reflect revisions proposed by the applicant. 

11. Revised Figure 1.2-12, “Reactor Building, Arrangement Plan at Elevation 31700/38200 mm” 

to reflect revisions proposed by the applicant. 

12. Revised Section 1.4, “Identification of Agents and Contractors,” to add a subsection 

number (1.4.1) and title to the existing discussion and add new subsections 1.4.2, “Applicant 

- Aircraft Impact Rule Amendment to Design Certification,” and 1.4.3, “Other Contractors 

and Participants.” 

13. Revised Section 1.8.3, “Applicability of Experience Information,” to add AFI to the list of new 

and novel design features used in the ABWR. 

14. Revised DCD Table 1.9-1, “Summary of ABWR Standard Plant COL License Information,” to 

add the following COL license information items: 

ITEM NO SUBJECT

9.39 Power Supply for Alternate Feedwater Injection Equipment

9.40 Test and Surveillance Intervals for Alternate Feedwater Injection Equipment,

19.9k Procedures for Use of Alternate Feedwater Injection

19.9l Procedures to Depressurize the Reactor Pressure Vessel (RPV) from the AFI 

Pump House

19.9m Verification of Environmental Conditions in the AFI Pump House

19.9n Description of Electrical Power Supply for AFI Equipment.

In the DCD amendment dated May 12, 2010 (ML101340548), the applicant deleted the COL 

license information items listed above. 

1.3 Regulatory Basis 

The regulatory basis and acceptance criteria for reviewing the introduction and general 

description of plant are in Section 1.0 of NUREG–0800, “Standard Review Plan for the Review 

of Safety Analysis Reports for Nuclear Power Plants,” Revision 3 (SRP). 

In particular, the applicable regulatory requirements for the ABWR DCD Amendment for Aircraft 

Impact Assessment (AIA) are as follows: 

• 10 CFR 52.47 requires the technical application information. 

• 10 CFR 52.47(a)(7) requires the applicant to demonstrate it is technically qualified to 

engage in the proposed activities. 
1-3

1.4 Technical Evaluation 

NRC staff reviewed the applicant’s proposed revision to the certified ABWR DCD Tier 2. The 

staff reviewed the changes to DCD Chapter 1 and determined that they are necessary and 

adequately reflect the changes to the certified ABWR design requested by STPNOC. 

1.4.1 Format and Content (Section 1.1.1)

This section of the ABWR DCD describes the format and content of the DCD. The applicant 

added a sentence that states, “The STPNOC response to the aircraft impact rule is provided in 

Tier 2, Appendix 19S.” The staff confirmed that this is an accurate description of the information 

STPNOC is adding to the ABWR DCD and found the applicant’s proposed change acceptable. 

1.4.2 Design Process (Section 1.1.4)

This section of the ABWR DCD describes the process used by the original applicant in 

designing the ABWR. STPNOC added a sentence that states, “The design process for the 

STPNOC response to the aircraft impact rule is fully described in the STP 3 and 4 Quality 

Assurance Program Description, which is provided in Tier 2, Subsection 17.1.19.” The staff 

confirmed that the referenced documents contain an accurate description of the design process 

used by STPNOC and found the applicant’s proposed change to DCD Section 1.1.4 acceptable. 

1.4.3 General Plant Description (Section 1.2)

This section of the ABWR DCD Section 1.2.2, “Plant Description,” provides general descriptions 

of the plant structures and systems. STPNOC added Subsection 1.2.2.12.23, “Alternate 

Feedwater Injection (AFI) System,” to reference new Section 9.5.14, where the AFI system is 

described. STPNOC also added Subsection 1.2.2.16.16, “Alternate Feedwater Injection (AFI) 

Pump House,” to describe the location of the AFI Pump House and the equipment housed in the 

AFI pump house. The staff confirmed that these are appropriate additions to reflect the design 

changes to the ABWR DCD and found the applicant’s proposed changes acceptable. 

1.4.4 Figures

STPNOC revised several figures in Chapter 1 of the ABWR DCD, as listed in Section 1.2 of this 

SER, to reflect revisions proposed by the applicant. The staff confirmed that these figures 

accurately reflect changes to the ABWR DCD and found the applicant’s proposed changes 

acceptable. 

1.4.5 Identification of Agents and Contractors (Section 1.4)

This section of the ABWR DCD describes the qualifications of the applicant. STPNOC did 

not propose any changes to this section in its original application. The staff issued 

RAI 01- 2 requesting STPNOC to provide additional information demonstrating that 

STPNOC is technically qualified to supply the amended ABWR design in accordance with 

10 CFR 52.47(a)(7). The applicant’s response to RAI 01-2 dated February 8, 2010 

(U7-C-STP-NRC-100037, ML100470589), states that STPNOC is responsible for licensing, 

operating, maintaining, modifying, decontaminating, and decommissioning STP Units 1 and 2. 

STPNOC has fulfilled this role since 1997, and has had extensive experience with the design of 

nuclear structures, systems and components (SSCs). The applicant adds that STPNOC is 

responsible for the licensing and development of STP Units 3 and 4, including the detailed 
1-4

design of these two planned ABWR units. The applicant further states that its activities as a 

supplier for the amended portion of the ABWR design certification will be subject to the same 

controls as those for the STP Units 3 and 4 licensing, design, construction/pre-operation and 

operations activities affecting the quality and performance of safety-related nuclear plant SSCs 

and certain activities that are not safety-related but support safe plant operations, or where NRC 

guidance establishes program requirements. 

The applicant stated that STP Units 3 and 4 Quality Assurance (QA) Program provides for 

control of those activities (as indicated in the DCD Tier 2, Section 17.0) and that appropriate 

controls ensure compliance with 10 CFR Part 21 during design. The applicant indicated that 

these requirements are passed down to contractors via procurement documents as appropriate, 

and vendors are qualified in accordance the QA Program. 

The applicant stated that STPNOC has entered into an engineering, procurement and 

construction (EPC) contract with Toshiba America Nuclear Energy Corporation (TANE), a 

Delaware corporation, which is a subsidiary of Toshiba. TANE's obligations include supply of 

the certified ABWR design for STP Units 3 and 4. TANE is the overall Project Manager, and is 

performing engineering and other responsibilities. TANE is being assisted by Westinghouse 

Electric (Westinghouse), which is providing engineering for selected systems, primarily fuel and 

safety analyses and instrumentation and control (I&C) systems. The applicant stated that 

STPNOC will be responsible as a supplier for the scope of the amendment to comply with 

10 CFR 50.150. STPNOC expects to receive support from TANE and Westinghouse under the 

contracts described above, or under future agreements, as appropriate. 

Specifically, with respect to the proposed amendment to the ABWR design certification to 

comply with the 10 CFR 50.150, the applicant indicated that Westinghouse performed the 

analyses in accordance with the Nuclear Energy Institute (NEI) 07-13, “Methodology for 

Performing Aircraft Impact Assessments for New Plant Designs” with assistance from ERIN 

Engineering & Research, Inc. (ERIN). The applicant stated that ERIN has substantial expertise 

and familiarity with the methodology described in NEI 07-13, as demonstrated by its significant 

role on behalf of NEI in the preparation of NEI 07-13. Additionally, Westinghouse has 

performed this scope of work for another design certification applicant. Sargent & Lundy is 

responsible for the structural design. The applicant also stated that Sargent & Lundy is an 

experienced nuclear architect engineer and has extensive experience in the design of 

structures, systems and components for nuclear power plants. 

That applicant stated that, in connection with the EPC contract with TANE, it evaluated the 

capability of Toshiba to complete the STP Units 3 and 4 ABWR project to NRC standards, and 

in particular, its qualifications to supply the certified ABWR design in accordance with the 

requirements of 10 CFR 52.73. STPNOC provided the NRC with a report of the results of that 

evaluation, describing in detail the bases on which STPNOC determined that Toshiba is 

qualified to supply the certified ABWR design. Letter from Gregory T. Gibson to NRC re: "South 

Texas Project, Units 3 and 4, Submittal of Due Diligence Assessment of Toshiba Corporation's 

Qualification to Supply the Design of the Advanced Boiling Water Reactor," August 19, 2008 

(ML082350160). The NRC conducted an inspection to independently assess the basis upon 

which STPNOC determined that Toshiba is capable of providing the certified ABWR for STP 

Units 3 and 4. Based on the inspection, the NRC concluded that STPNOC had adequately 

demonstrated Toshiba's qualification to supply the certified U.S. ABWR for STP Units 3 and 4. 

Letter from John A. Nakoski to Mark A. McBurnett, STPNOC, re: "NRC Inspection 

Report 05200012/2009- 202 AND 5200013/2009-202," August 28, 2009 (ML092370709). 
1-5

Subsequent to submitting its RAI response, STPNOC revised its application and added a 

subsection number (1.4.1) and title to the existing discussion on the original applicant and add 

new subsections 1.4.2, “Applicant - Aircraft Impact Rule Amendment to Design Certification,” 

and 1.4.3, “Other Contractors and Participants.” The new subsections describe STPNOC’s 

qualifications and those of its primary contractors and subcontractors that supported the 

amendment to the ABWR design certification. The staff’s review of the information added to this 

section determined that the applicant has adequately described its qualifications and those of its 

primary contractors and subcontracts and the staff finds the applicant’s proposed changes 

acceptable. 

To further evaluate STPNOC’s capabilities as a prospective design supplier, the NRC 

performed an inspection of the applicant’s quality controls over the work done by its contractors 

in support of its request to amend the ABWR design certification to comply with 10 CFR 50.150. 

The staff’s evaluation of information obtained during that inspection is discussed in 

Section 17.1.19 of this SER. 

The staff evaluated the information provided by the applicant in its response to RAI 01-2, its 

revisions to Chapter 1 of the ABWR DCD, and the results of the staff’s inspection of the 

applicant’s quality controls. As stated in the applicant’s RAI response, the staff has previously 

concluded that STPNOC had adequately demonstrated Toshiba's qualification to supply the 

certified U.S. ABWR for STP Units 3 and 4. Based on the previous determination and the 

additional information provided by STPNOC in the response to RAI 01-2, the staff determined 

that STPNOC and its contractors are technically qualified to perform the design work associated 

with the amended portion of the ABWR design represented by STPNOC’s application and to 

supply the amended portion of the ABWR design. However, based on the staff’s evaluation of 

the information provided in the applicant’s RAI response and reviewed during the staff’s quality 

controls inspection the staff determined that STPNOC, by itself, is not technically qualified to 

supply the amended portion of the ABWR design certification represented in STPNOC’s DCD, 

Revision 3, However, the staff determined that STPNOC and TANE acting together are 

qualified to supply the amended portion of the ABWR design certification represented in 

STPNOC’s DCD amendment, Revision 3. Therefore, the staff intends to propose inclusion of 

language to that effect in any rule to certify STPNOC’s proposed amendment to the ABWR 

design certification to ensure that the basis for any NRC finding of technical qualification in 

support of this design certification amendment remains valid. 

1.4.6 Applicability of Experience Information (Section 1.8.3)

This section of the ABWR DCD describes the applicant’s evaluation of nuclear field experience. 

This section also contains a list of the new and novel design features used in the ABWR 

STPNOC added the AFI system to the list of new and novel design features used in the ABWR. 

The staff’s review determined that this addition accurately reflects the new and novel design 

features that STPNOC is proposing to add to the ABWR DCD and the staff finds the applicant’s 

proposed changes acceptable. 

1.5 Conclusion 

NRC staff reviewed the general description of the plant design aspects of the STPNOC's 

application to amend the Design Certification Rule for the ABWR. The staff’s review of the 

changes to DCD Chapter 1 determined that they are necessary and adequately reflect the 

changes to the certified ABWR design requested by STPNOC. The staff concluded that the 

applicant has adequately addressed the requirements of 10 CFR 52.47(a)(7). However, the 
1-6

staff‘s determination is based on the condition that STPNOC and TANE acting together will 

supply the amended portion of the ABWR design to any future license applicants. The staff will 

include language to that effect in any rule to certify STPNOC’s proposed amendment to the 

ABWR design certification. The staff reviewed the proposed changes to DCD Tier 2, Chapter 1 

and found them acceptable. The staff determined that there would be no adverse impacts from 

complying with the requirements for consideration of aircraft impacts on conclusions reached by 

the NRC in its review of the original ABWR design certification (NUREG-1503). 
3-1

3 DESIGN OF STRUCTURES, COMPONENTS, 

EQUIPMENT, AND SYSTEMS 

3.2.1 Seismic Classification 

3.2.1.1 Introduction 

As part of the application, STPNOC proposes to add a new AFI system and pump house. This 

section of the SER discusses the staff’s evaluation of the seismic classifications of the proposed 

additions and the technical bases for the staff’s acceptance of these changes. Also, this section 

discusses the staff’s determination of whether the changes adversely impact conclusions 

reached by the NRC in its review of the original ABWR design certification. 

3.2.1.2 Summary of Application 

DCD Table 3.2-1 and Figure 9.5-6 show the classification of the new AFI system beyond the 

first check valve from the feedwater system as Non-Safety Class, which is equivalent to 

non-nuclear safety-related or non-safety-related. Table 3.2-1 also shows both the AFI System 

and the AFI pump house as Safety Class N and, consistent with the methodology applied in the 

certified design, the seismic classification of Safety Class N SSCs are identified as not 

applicable and equivalent to a non-Seismic Category I classification. The amended ABWR 

DCD Revision 3, Table 3.2-1 shows the instrumentation piping and supports forming part of the 

containment boundary and a portion of the piping beyond the containment outermost isolation 

valves as Seismic Category I. 

3.2.1.3 Regulatory Basis 

The regulatory basis and acceptance criteria for reviewing the seismic classification are in 

Section 3.2.1 of NUREG–0800. The guidance in the Section 3.2.1 of NUREG–0800, “Seismic 

Classification,” references Regulatory Guide (RG) 1.29 for seismic classification of various 

SSCs. General Design Criterion (GDC) 2, with specific guidance included in RG 1.29 for the 

seismic classification applies to the extent the changes could impact conclusions that were 

previously made. 

3.2.1.4 Technical Evaluation 

Staff reviewed the seismic classification of the additional structures, systems and components 

according to the guidance in SRP 3.2.1. The NRC staff’s review determined that the NRC’s 

requirements that apply to the design, construction, testing, operation and maintenance of 

design features and functional capabilities for design-basis events do not apply to design 

features or functional capabilities selected by the applicant solely to meet the requirements of 

the final AIA rule. The staff reviewed the seismic classification of the additional SSCs and 

determined that the non-safety-related SSCs that are exempt from the seismic classification 

guidance need not be Seismic Category I, provided that they are evaluated for adverse systems 

interactions consistent with the certified DCD. As identified in SRP 3.7.2 and SRP 3.12, seismic 

interactions for structures and piping are evaluated and reviewed in other sections of the DCD 

and SER. Subsection 3.7.2.8 of the ABWR DCD describes the criteria applied to evaluate the 

interaction of non-seismic Category I SSCs with Seismic Category I SSCs. DCD Note f to 

Table 3.2-1 also identifies that equipment that is not safety-related but could damage Seismic 
3-2

Category I equipment if its structural integrity failed is verified analytically and designed to 

assure its integrity under seismic loading resulting from the safe-shutdown earthquake (SSE). 

Therefore, the staff determined that classification of the non-safety-related Safety Class N 

portion of the AFI system and AFI Pump House as not Seismic Category I is acceptable and 

consistent with the AIA rule, since the non-safety-related portions of the AFI system and the AFI 

Pump House are selected solely to meet requirements of the rule and there are criteria 

established to evaluate adverse systems interactions. The staff’s review of the initial 

Amendment also determined that the nitrogen supply line, nitrogen bottle, and AFI system 

outside of the pump house were not included in Table 3.2-1 and requested additional 

information to clarify the classification of these items. NRC staff issued RAI 03.02.02-1 

requesting the applicant to review the SSCs being added to the AIA and update classification 

Table 3.2-1 to include SSCs shown on the piping and instrumentation diagrams. 

The applicant’s response to RAI 03.02.02-1 dated March 3, 2010, (U7-C-STP-NRC-100056, 

ML100640162) states that the additional nitrogen gas supply system is not required to meet the 

AIA rule and the system is being removed from the DCD amendment application. The response 

adds that the AFI instrument lines and instruments will be included in Table 3.2-1 and states 

that they are safety-related with the same classification as the existing instrumentation to which 

it is connected. The amended ABWR DCD, Revision 3 includes these changes and shows the 

instrumentation piping and supports forming part of the containment boundary as Seismic 

Category I. Classification of these safety-related items as Seismic Category I is consistent 

with RG 1.29 and GDC 2 and is therefore acceptable. During its review, the staff also 

determined that the portion of the AFI system outside of the AFI pump house was not identified 

in Table 3.2-1. DCD Table 3.2-1 has been revised (ML101190120) to include the 

non-safety-related AFI piping outside the pump house as non-seismic. Therefore, the staff 

found the applicant’s RAI responses and revised DCD AIA Amendment Table 3.2-1 acceptable. 

Therefore, RAI 03.02.02-1 is resolved and closed. 

Consistent with SRP 3.2.1, electrical items are not within scope of the 3.2.1 review and 

typically their seismic classification is addressed in Chapter 8. Since Chapter 8 of the 

application was not revised, the seismic classification of electrical items to support the AFI is 

evaluated in this subsection. DCD subsection 9.5.14.1 states that the power supply for the 

pump and motor-operated valves is a non-safety-related power supply and independent of the 

emergency power supplies and meets the requirements of non-Class 1E power, as described in 

Chapter 3. Staff concurs that the non-safety-related power supply need not be Class 1E or 

classified as Seismic Category I and is acceptable in regard to systems interactions, since it is 

independent of the emergency power supplies. 

3.2.1.5 Conclusion

The staff reviewed the seismic classification aspects of the STPNOC's application to amend the 

Design Certification Rule for the ABWR. The staff’s review focused on the determination of the 

appropriate seismic classification for the new AFI system and pump house. The staff found that 

the applicant has adequately demonstrated that the proposed additions of the AFI pump house, 

AFI system piping, supports and power supply, as described in the application to amend the 

ABWR DCD, comply with RG 1.29, GDC 2 and the AIA rule relative to seismic classifications. 

The staff also found that these seismic classifications do not alter the fundamental safety 

decisions in the original ABWR design certification (NUREG–1503). 
3-3

3.2.2 Quality Group Classifications 

3.2.2.1 Introduction

As part of the application, STPNOC proposes to add a new AFI system and pump house. This 

section of the safety evaluation discusses the staff’s evaluation of the quality group 

classifications of the proposed additions and the technical basis for the staff’s acceptance of 

these quality group classifications. Also, this section discusses the staff’s determination of 

whether the changes adversely impact conclusions reached by the NRC in its review of the 

original ABWR design certification. 

3.2.2.2 Summary of Application

The amended ABWR DCD Table 3.2-1 and Figure 9.5-6 show the classification of the new AFI 

System beyond the first check valve from the feedwater system as Non-Safety Class and Safety 

Class N, respectively, which are equivalent to non-nuclear safety-related or non-safety-related. 

Noted to Table 3.2-1 identifies that quality group is not applicable to the non-safety-related AFI 

system and the classification of Safety Class N SSCs in the DCD are typically identified with no 

quality group classification. In the amended ABWR DCD Revision 3, Table 3.2-1 also shows 

the instrumentation piping and supports forming part of the containment boundary and a portion 

of the piping beyond the containment outermost isolation valves as Safety Class 2 and Quality 

Group B. In the initial Amendment, DCD Figure 6.7-1 showed a modified high-pressure 

nitrogen gas supply line. 

3.2.2.3 Regulatory Basis

The regulatory basis and acceptance criteria for reviewing the quality group classifications are in 

Section 3.2.2 of NUREG–0800. The guidance in Section 3.2.2 of NUREG–0800, “Quality Group 

Classification,” references RG 1.26 for quality group classification of various SSCs. GDC 1, 

with specific guidance included in RG 1.26 for the quality group classification applies to the 

extent the changes could impact conclusions that were previously made. 

3.2.2.4 Technical Evaluation

NRC staff reviewed the quality group classification of the additional systems and components 

according to the guidance in SRP 3.2.2. Consistent with SRP 3.2.2, quality group classification 

applies to pressure boundary items and their supports but not to structures. The staff’s review 

determined that the NRC requirements that apply to the design, construction, testing, operation 

and maintenance of design features and functional capabilities for design-basis events do not 

apply to design features or functional capabilities selected by the applicant solely to meet the 

requirements of the final AIA rule. The staff determined that SSCs that are exempt from the 

quality group classification guidance need not be designated with a specific quality group 

provided that they have appropriate QA requirements applied and are evaluated for adverse 

system interactions. DCD Note e to Table 3.2-1 identifies that for equipment such as the AFI 

system that is not safety-related, elements of 10 CFR Part 50, Appendix B are generally 

applied, commensurate with the importance of the equipment’s function. Therefore, the staff 

finds that appropriate elements of 10 CFR Part 50 are applied to the AFI system and 

classification of the Safety Class N AFI system with no quality group designation is acceptable 

and consistent with the certified DCD and the AIA rule. The staff’s review also determined that 

the nitrogen supply line, nitrogen bottle, and AFI system outside of the pump house are not 

included in Table 3.2-1. In RAI 03.02.02-1, staff requested the applicant to provide additional 
3-4

information to clarify the classification of these items. The staff also asked the applicant to 

review the SSCs being added in the AIA and to update classification Table 3.2-1 to include the 

SSCs shown on the piping and instrumentation diagrams (P&IDs). 

The applicant’s response to RAI 03.02.02-1 dated March 3, 2010 (U7-C-STP-NRC-100056, 

ML100640162) states that the additional nitrogen gas supply system is not required to meet the 

AIA rule and the system is being removed from the DCD amendment application. The response 

further identifies that the AFI instrument lines and instruments will be added to Table 3.2-1 and 

that they are safety-related with the same classification as the existing instrumentation to which 

they are connected. The revised DCD Table 3.2-1 designates the AFI instrument lines and 

instruments as Quality Group B. Classification of these safety-related instrumentation items as 

Quality Group B is consistent with RG 1.26 and GDC 1 and therefore acceptable. The staff 

review also determined that the portion of the AFI system outside of the AFI pump house is not 

addressed and the revised Table 3.2-1 submitted with the RAI response only shows the AFI 

system located in the AFI pump house. The applicant subsequently revised DCD Table 3.2-1 

(ML101190120) to include the non-safety-related AFI piping outside the pump house with no 

quality group classification, consistent with the methodology in the existing DCD. The applicant 

also deleted the nitrogen supply modifications. This revised DCD is consistent with the AIA rule, 

because this non-safety-related portion of the AFI system serves no safety function. Therefore, 

the staff found the applicant’s revised DCD AIA Amendment Table 3.2-1 acceptable, and 

RAI 03.02.02-1 is resolved and closed. 

3.2.2.5 Conclusion

The staff reviewed the quality group classifications aspects of the STPNOC's application to 

amend the Design Certification Rule for the ABWR. The staff’s review focused on the 

determination of the appropriate quality group classification for pressure boundary items and 

their supports in the AFI system. The staff found that the applicant has adequately 

demonstrated that the quality group classification of the AFI system, as described in the 

application to amend the ABWR DCD, comply with RG 1.26, GDC 1 and the AIA rule relative to 

quality group classifications. The staff also found that these changes do not change the 

fundamental safety decisions in the original ABWR design certification (NUREG–1503). 

3.8.6 Alternate Feedwater Injection Pump House 

3.8.6.1 Introduction

As part of the application, STPNOC proposes to add an AFI pump house to provide enclosure 

and support to a new AFI system. In addition, STPNOC proposes to add a series of fire doors 

to Seismic Category I structures. This section of the safety evaluation discusses the staff’s 

evaluation of the proposed additions and the technical basis for the staff’s acceptance of these 

proposed additions. Also, this section discusses the staff’s determination of whether the 

additions adversely impact conclusions reached by the NRC in its review of the original ABWR 

design certification. 

3.8.6.2 Summary of Application

The applicant proposes to add or change several DCD sections including DCD Tier 2, 

Revision 3, Subsections 1.2.2.12.23, “AFI system”; 1.2.2.16.15, “AFI Pump House”; part of 

Table 1.9-1; part of Table 3.2-1 and the fire doors discussed in DCD Tier 2, Revision 3, 

Section 9A.4.1. 
3-5

3.8.6.3 Regulatory Basis

The regulatory basis and acceptance criteria for reviewing AFI pump house are in Sections 3.7 

and 3.8 of NUREG–0800. In particular, the staff’s review is based on 10 CFR Part 50, 

Appendix A, GDC 2 and 4 and applicable provisions of SRP Sections 3.3.1, 3.3.2, 3.4.2, 3.5.3, 

3.7 and 3.8. 

3.8.6.4 Technical Evaluation

Alternate Feedwater Injection Pump House 

The applicant proposes to add a new AFI pump house with a non-safety-related AFI system 

located outside of the reactor building. Subsection 1.2.2.16.15 of the amended ABWR DCD 

states that the location of the AFI pump house is a remote distance from the reactor building. 

The amended Table 3.2-1, “Classification Summary,” of the ABWR DCD classifies the AFI 

pump house as a non-safety class structure; so, the seismic design requirements for Seismic 

Category I structures in SRP Sections 3.7 and 3.8 for the SSE are not applicable. The staff 

agreed with the applicant's statement that the seismic design requirements provided in the SRP 

Sections 3.7 and 3.8 are not applicable to the AFI pump house because of its non-seismic 

Category I designation. 

The staff considered the potential of Seismic II/I interaction effects between the AFI pump house 

and it’s adjacent Seismic Category I structures. Because the AFI pump house will be located at 

least 300 ft away from Seismic Category I structures, the staff agreed that the seismic II/I 

interaction issue is not applicable to the AFI pump house. 

Additionally, the staff evaluated potential impacts, or changes to the structural and seismic 

design bases of Seismic Category I SSCs as described in the ABWR DCD that might result 

from the addition of the AFI pump house. By being located at least 300 ft away from the 

Seismic Category I structures, the staff concluded that the original structural and seismic design 

bases of the ABWR Seismic Category I SSCs will remain intact and unaffected by the addition 

of the AFI pump house. Based on the above evaluation, the staff concluded the addition of the 

AFI pump house as proposed in the amended ABWR DCD is acceptable. 

Replacements and Additions of Fire/Pressure Rated Doors to Seismic Category I Structures

The staff reviewed information related to fire doors presented in Section 9A.4 of the amended 

ABWR DCD. The review indicated that the applicant proposes to upgrade fire doors as 

described in Section 9A.4 and the Figures in Sections 1.2 and 9A as part of the fire barriers 

qualification program. The proposed upgrade involves replacements of nonrated, 3-hour, fireresistant or non-fire-rated doors with 5-psid door or two 3-hour-rated fire doors for the walls 

surrounding the above listed fire areas. 

Based on the review, the staff issued RAI 03.08.04-4 and requested the applicant to discuss 

and confirm that the above noted changes and additions of doors as well as their resulting 

structural configuration changes were evaluated (including wall structural integrity analyses, as 

needed) to ensure that the changes are bounded within their original structural design basis, 

and the affected walls will continue to maintain their structural integrity and perform their 

intended safety functions. 
3-6

The applicant’s response to RAI 03.08.04-4 dated February 25, 2010 (U7-C-STP-NRC-100040, 

ML100600410) states that the overall structural characteristics of the reactor building are 

unchanged by replacing these doors. Because the calculations for room pressurization already 

account for the presence of the doors, the change to two doors or to a pressure-rated door 

represents a very small change in the analysis. Consequently, replacement of the doors is 

expected to have a minimal effect on the overall structural performance of the reactor building. 

The applicant adds that the detailed structural analysis must be performed by the COL Holder 

following completion of the detailed design which will include such details as reactor building 

internal wall location, wall dimensions, wall materials, the replacement of the 3-hour fire 

resistant or non-rated fire doors with 5-psid or two 3-hour-rated fire doors, and any other 

changes resulting from the final fire hazards analyses. Further, the interior wall structural 

analysis will be performed as part of the complete reactor building design in accordance with 

ABWR DCD Tier 1, Section 2.15.10; the associated inspections, tests, analyses, and 

acceptance criteria (ITAAC) in DCD Tier 1, Table 2.15.10; and Tier 2, Appendix 3H.1. 

The staff reviewed the above applicant’s response and determined that the response was 

incomplete and needed to be augmented in order to be acceptable. The staff requested that the 

applicant provide an ITAAC covering the proposed replacements and additions of fire/pressure 

rated doors to address the reconciliation of the design basis loads as a result of as-built 

conditions. The above staff’s position was provided to the applicant on May 5, 2010. 

In a letter dated May 27, 2010 (U7-C-STP-NRC-100117, Attachment 2, ML101530610), the 

applicant provided a revised response to RAI 03.08.04-4 stating that: 

The detailed structural analysis must be performed following completion of 

detailed design, which will include such details as reactor building internal wall 

location, wall dimensions, wall materials, the replacement of the 3-hour fire 

resistant or non-rated fire doors with 5-psid or two 3-hour rated fire doors, and 

any other changes resulting from the final fire hazards analyses. The interior wall 

structural analysis will be performed as part of the complete reactor building 

design in accordance with ABWR DCD Tier 1, Subsection 2.15.10, the 

associated design ITAAC provided in Table 2.15.10, Item 10, and Tier 2, 

Appendix 3H.1. As identified above, the changes and additions of fire doors, as 

identified in the Appendix 9A markups for the DCD amendment application, are 

within the scope of these existing analysis requirements. These requirements 

ensure that the changes and additions of fire doors are considered in the 

structural design basis and evaluation of the design basis loads in accordance 

with ABWR DCD Tier 1, Subsection 2.15.10. 

The staff evaluated the above revised response to RAI 03.08.04-4 and concluded that the 

replacements and additions of fire/pressure rated doors will be included in the original structural 

design basis as referenced in DCD Tier 1, Section 2.15.10, and the implementation of the 

associated ITAAC design commitments within DCD Tier 1, Table 2.15.10, Item 10. Therefore, 

the staff has reasonable assurance that the certified design will not be negatively impacted by 

replacements and additions of the fire/pressure rated doors to Seismic Category I structures. 

Based on this finding, RAI 03.08.04-4 is considered resolved and closed. 
3-7

 Impact of the Addition of the AFI System on the Adequacy of the Seismic Design Basis of the 

Original Feedwater Piping and Supports

The staff reviewed the information related to the proposed addition of the AFI system, including 

the amended ABWR DCD Figure 9.5-6, “Alternate Feedwater Injection System Schematic.” 

The staff’s review assessed the potential impact on the adequacy of the seismic design basis 

of the original feedwater piping and supports. Based on the review, the staff issued 

RAI 03.08.04-2 requesting that the applicant confirm whether a piping seismic response 

analysis of the modified reactor water cleanup system (CUW) tie-in lines to the feedwater 

system, modeled in conjunction with the AFI line, was performed to ensure that the effects of 

the AFI line addition are accounted for within the seismic design basis of the original feedwater 

piping and supports. Also, the staff asked the applicant to clarify whether the design basis of 

the feedwater system will be adversely affected by the addition of the AFI line. Finally, the staff 

asked the applicant to summarize the results of these analyses, if any. 

The applicant’s response to RAI 03.08.04-2 dated February 25, 2010 (ML100600410) states 

that the piping seismic response analysis of the modified CUW tie-in line to the feedwater 

system has not yet been performed. The applicant adds that this analysis must be performed 

by the COL Holder following completion of detailed design and after determination of such 

details such as pipe routing, location of pipe supports and restraints, and final line sizes. In 

addition, the final piping seismic analysis of the feedwater system will account for the effects of 

attached non-safety-related piping systems, including the CUW water system and the AFI 

piping. Furthermore, this piping seismic analysis will be performed according to the 

requirements identified in the ABWR DCD Tier, Section 3.3, the associated ITAAC provided in 

Table 3.3, and the seismic analysis methods identified in Tier 2, Section 3.7.3. Because the 

tie-in of the AFI line is to the non-safety-related portion of the CUW lines, and consequently 

does not tie-in directly to a safety-related line, the applicant expects that the design basis of the 

original feedwater piping and supports will not be adversely affected by this added AFI tie-in 

line. 

The staff reviewed the above applicant’s response to RAI 03.08.04-2 and determined that the 

response was incomplete and needed to be augmented in order to be acceptable. The staff 

requested that the applicant provide an ITAAC covering the proposed AFI tie-in line to address 

the reconciliation of the design basis loads as a result of the as-built conditions. The above 

staff’s position was provided to the applicant on May 5, 2010. 

In a letter dated May 27, 2010 (U7-C-STP-NRC-100117, Attachment 1, ML101530610), the 

applicant provided a revised response to RAI 03.08.04-2 stating that: 

The piping seismic response analysis of the modified reactor water cleanup 

system (CUW) tie-in lines to the feedwater system has not yet been performed. 

This analysis must be performed following completion of detailed design after 

determination of such details as pipe routing, location of pipe supports and 

restraints, final line sizes, etc. The final piping seismic analysis of the feedwater 

system will account for the effects of attached non-safety piping systems, 

including the CUW water system and the AFI piping. This piping seismic 

analysis is performed according to the requirements identified in the ABWR DCD 

Tier 1, Section 3.3, the associated design ITAAC provided in Table 3.3, Items 1 

through 3, and the seismic analysis methods identified in Tier 2, Section 3.7.3. 

As identified above, adding AFI piping and tie-in to the CUW system is within the 

scope of these existing analysis requirements for the piping design ITAAC. 
3-8

These requirements ensure that any effects of the non-safety-related AFI piping 

on ASME Class 1, 2 and 3 piping are considered. Because the tie-in of the AFI 

line is to the non-safety-related portion of the CUW lines, and consequently does 

not tie-in directly to a safety-related line, it is expected that the design basis of 

the original feedwater piping and supports will not be adversely affected by this 

added AFI tie-in line. 

The staff evaluated the above revised response to RAI 03.08.04-2. The staff verified that the 

plant-specific piping layout and support configuration data that show the tie-in of the AFI line is 

to the non-safety-related portion of the CUW lines and does not tie-in directly to a safety-related 

line. The staff concluded that the piping seismic response analysis of the modified CUW tie-in 

lines to the feedwater system will be accounted for within the seismic design basis of the 

original feedwater piping and supports. Therefore, the staff has reasonable assurance that the 

certified design will not be negatively impacted by the addition of the AFI tie-in line. Based on 

this finding, RAI 03.08.04-2 is considered resolved and closed. 

3.8.6.5 Conclusion

NRC staff reviewed the structural and seismic design aspects of the STPNOC's application to 

amend the Design Certification Rule for the ABWR. Specifically, the applicant proposes to add 

an AFI pump house, a new AFI System, a series of fire doors, and other related components. 

The staff’s review focused on the determination of structural and seismic safety and design 

adequacy of the (1) proposed addition of an AFI pump house, (2) proposed replacements and 

additions of fire- and pressure-rated doors to Seismic Category I structures, and (3) impact 

assessment of the addition of the AFI System on the adequacy of the seismic design basis of 

the original feedwater piping and supports. The staff determined that there would be no 

adverse impacts from complying with the requirements for consideration of aircraft impacts on 

conclusions reached by the NRC in its review of the original ABWR design certification 

(NUREG-1503). 

The applicant also provided reasonable assurance that the original structural and seismic 

design bases included in the ABWR DCD for Seismic Category I SSCs will remain intact and 

unaffected by the addition of the AFI pump house and replacements and additions of 

fire/pressure rated doors to Seismic Category I SSCs. 

The staff also found that the proposed modifications do not change the fundamental safety 

decisions in the original FSER. Therefore, the structural and seismic aspects of the 

proposed amendment to the DCD are acceptable. 
5-1

5 REACTOR COOLANT SYSTEM AND 

CONNECTED SYSTEMS 

As part of the application, STPNOC revises DCD Figure 5.1-3, “Nuclear Boiler System P & ID, 

(Sheet 4 of 11),” to show the AFI system connection to the Nuclear Boiler System through the 

non-safety-related portion of the CUW tie-in lines to the feedwater system. The staff found this 

change acceptable, and is evaluated in Section 9.5.14 of this SER. 

5.2.2 Overpressure Protection 

5.2.2.1 Introduction

As part of the application, STPNOC proposed to add a solenoid valve to a SRV for overpressure 

protection. The overpressure protection in the ABWR is provided using 18 SRVs, of which 8 are 

part of the pressure set point groups and mounted on the four main steam lines between the 

reactor vessel and the first isolation valve inside the containment. This section of the SER 

safety evaluation discusses the staff’s evaluation of the over pressure protection of the 

proposed changes and the technical basis for the staff’s acceptance of these proposed 

changes. Also, this section discusses the staff’s determination of whether the changes 

adversely impact conclusions reached by the NRC in its review of the original ABWR design 

certification. 

5.2.2.2 Summary of Application

In the initial submittal dated June 30, 2009 (ML092040048), the applicant proposed to add a 

paragraph to DCD Subsection 5.2.2.4.1 describing the addition of a solenoid valve to SRV–E 

and a nitrogen supply line to supply nitrogen from the AFI pump house. The intent of this 

modification was to open one SRV from the pump house after the aircraft impact. The 

applicant also proposed changes to the DCD Figures 5.1-3 “Nuclear Boiler System P & I D,” 

(Sheets 1, 2, 9 and 11 of 11) to incorporate the new solenoid valve for the mitigating functions 

from the pump house. 

However, in response to RAI 06.02.04-1, the applicant, in letter dated January 13, 2010 

(U7-C-STP-NRC-100009, ML100190088), withdrew the changes in the DCD Section 5. 

Therefore, there are no changes to the overpressure protection system, including the SRVs, 

due to the AIA Amendment. Also, in response to RAI 05.02.02-1, the applicant reiterates and 

references the response to RAI 06.02.04-1. Therefore, there are no changes to the 

overpressure protection system, including the SRVs, in regard to the AIA Amendment. In the 

revised DCD AIA Amendment dated May 12, 2010 (ML101340548), the applicant withdrew the 

changes. 

5.2.2.3 Regulatory Basis 

The regulatory basis for reviewing the overpressure protection is in Section 5.2.2 of 

NUREG-0800, Revision 3. 

In particular, the acceptance criteria are based on meeting 10 CFR Part 50, Appendix A, 

GDC 15 and 31. The acceptance criteria are based on GDC 15, as it relates to the reactor 

coolant system and associated auxiliary, control, and protection systems being designed with 
5-2

sufficient margin to ensure that the design conditions of the Reactor Coolant Pressure Boundary 

(RCPB) are not exceeded during any condition of normal operation including anticipated 

operating occurrences (AOOs). Because the amendment included an addition of a new 

solenoid valve to modify the operation of SRV-E, it is required that the safety valves have 

sufficient capacity to perform their intended safety-related function of limiting the pressure to 

less than 110 percent of the RCPB design pressure. 

SRP Section 5.2.2 states that the acceptance criteria are based on GDC 31, as it relates to the 

fracture behavior of the RCPB. Overpressure protection during low temperature operation is not 

considered for the ABWR or addressed because there is a very low probability that the ABWR 

will operate in water-solid conditions. Therefore, overpressure protection during low 

temperature conditions is not addressed for the ABWR. 

5.2.2.4 Technical Evaluation 

NRC staff reviewed the information in the DCD amendment. The staff issued RAI 05.02.02-1 

requesting the applicant to confirm that the addition of the new solenoid valve will not prevent 

the SRV-E from performing its intended safety-related function. The applicant’s response to 

RAI 05.02.02-1 dated January 20, 2010 (U7-C-STP-NRC-100026, ML100250139), states that 

the additional nitrogen gas supply system and the additional solenoid valve described in the 

initial DCD amendment submittal is not required to meet the AIA rule. The overpressure 

protection system is not affected by any of the design features incorporated into the design to 

meet the AIA rule. The SRVs will open and close as the reactor pressure increases and 

decreases. Emergency core cooling systems (ECCS) and the ADS will be available for core 

cooling. Core cooling is achieved without any modifications to the overpressure protection and 

ECCS. The staff’s evaluation of core cooling is described in SER Subsection 19.S.4.2, “Key 

Design Features for Core Cooling” and Subsection 19.S.4.3, “Key Design Features that Protect 

Core Cooling Design Features” of this SER. The staff reviewed the revised DCD AIA 

Amendment and confirmed that there are no changes to the certified DCD Section 5.2.2. 

Therefore, RAI 05.02.02.-1 is resolved and closed. 

In addition, the staff issued RAI 05.02.02-2 requesting the applicant to describe the type of 

analyses performed to ensure that the non-safety AFI line connected to the CUW tie-in lines to 

the feedwater system is designed so that it will not inadvertently impact the ability of 

safety-related RCIC system, which injects water into the reactor through the feedwater system, 

to perform its intended functions. This RAI is addressed in Subsection 9.5.14.4 of this SER 

because it directly relates to the AFI system. 

5.2.2.5 Conclusion 

NRC staff reviewed the overpressure protection aspects of the STPNOC's application to amend 

the Design Certification Rule for the ABWR. The staff’s review found the applicant has 

adequately addressed the applicable requirements. The staff determined that there would be 

no adverse impacts from complying with the requirements for consideration of aircraft impacts 

on conclusions reached by the NRC in its review of the original ABWR design certification 

(NUREG-1503).
6-1

6 ENGINEERED SAFETY FEATURES 

6.2.4 Containment Isolation System 

6.2.4.1 Introduction 

As part of the application, STPNOC proposes changes to the containment isolation information 

for high-pressure nitrogen gas supply system. This section of the ABWR DCD addresses the 

isolation systems including valves and associated piping that allow normal or emergency 

passage of fluids through the containment boundary while preserving the ability of the boundary 

to prevent or limit the escape of fission products from postulated accidents. This section of the 

SER discusses the staff’s evaluation of the containment isolation system of the proposed 

additions and the technical basis for the staff’s acceptance of these proposed changes. Also, 

this section discusses the staff’s determination of whether the changes adversely impact 

conclusions reached by the NRC in its review of the original ABWR design certification. 

6.2.4.2 Summary of Application 

In the initial submittal dated June 30, 2009 (ML092040048), the applicant proposed changes to 

DCD Section 6.7, “High Pressure Nitrogen Gas Supply System,” and to Section 6.2.4, 

“Containment Isolation System.” The applicant provided markups to the DCD to add an 

additional non-safety-related nitrogen gas storage bottle capable of supplying nitrogen to one of 

the non-ADS SRVs from the AFI pump house to allow system depressurization in the event of 

the loss of nitrogen supply in the reactor building. In the revised DCD AIA Amendment dated 

May 12, 2010 (ML101340548), the applicant removes the proposed changes to DCD 

Section 6.7. 

6.2.4.3 Regulatory Basis

The regulatory basis for reviewing containment isolation system is in Section 6.2.4 of 

NUREG-0800, Revision 3. 

6.2.4.4 Technical Evaluation 

The staff reviewed the proposed changes and issued RAI 06.02.04-1 requesting the applicant to 

provide additional details on the design and operation of added containment isolation valves 

P54-F301/F302 shown in Figure 6.7-1 and Table 6.2-7 and associated other changes in the 

DCD amendment markups. In response to this RAI in the letter dated January 13, 2010 

(U7-C-STP-NRC-100009, ML100190088), the applicant states that because the additional 

nitrogen gas supply system described in the initial DCD amendment submittal is not required to 

meet the AIA Amendment rule, this system is being removed from the DCD amendment. 

Consequently, all of the DCD amendment markups associated with this system will be deleted. 

As a result, there will no longer be a change to DCD Figure 6.7-1 and Table 6.2-7 and 

associated other changes of the DCD for this amendment. On May 12, 2010, the applicant 

submitted a revised DCD AIA amendment (ML101340548). The staff reviewed the revised DCD 

AIA Amendment, and confirmed that the applicant is not modifying the containment isolation 

system. Based on the applicant’s revised amendment, RAI 06.02-04-1 is considered resolved 

and closed. 
6-2

6.2.4.5 Conclusion

NRC staff reviewed the containment isolation system of the STPNOC's application to amend the 

Design Certification Rule for the ABWR. The staff’s review concluded that the containment 

isolation system is not modified to comply with the AIA rule and remains unchanged as certified 

originally in NUREG–1503. 
7-1

7 INSTRUMENTATION AND CONTROLS

7.7 Instrumentation and Controls not Required for Safety 

7.7.1 Introduction 

As part of the application, STPNOC proposes to add additional instrumentation (reactor water 

level, suppression pool water level, wetwell pressure, and reactor vessel pressure) in the AFI 

pump house. The information in this chapter emphasizes those instruments and associated 

equipment that are applicable to the ABWR design amendment of the ABWR DCD that 

addresses the AIA rule. This section of the SER discusses the staff’s evaluation of the I&C 

required for the proposed additions and the technical basis for the staff’s acceptance of these 

proposed changes. Also, this section discusses the staff’s determination of whether the 

changes adversely impact conclusions reached by the NRC in its review of the original ABWR 

design certification. 

7.7.2 Summary of Application 

Assessment of the AIA rule on the ABWR reactor design was performed by the applicant in 

accordance with 10 CFR 50.150(a) to identify and incorporate into the design those design 

features and functional capabilities to show that with the reduced use of operator actions, (i) the 

reactor core remains cooled or the containment remains intact; and (ii) spent fuel cooling or 

spent fuel pool integrity is maintained. In Section 7.7, “Control Systems Not Required for 

Safety,” the DCD amendment addresses additional features of the design to comply with the 

NRC AIA rule. 

The applicant identifies a need for an AFI system that is included to conform to the AIA rule and 

is located in a remote facility. Section 7.7 of the DCD amendment references Section 9.5.14, 

which contains the details of the AFI system I&C. The AFI system is designed to mitigate the 

consequences of an aircraft impact. 

The AFI system is capable of injecting water (≥3028 Lpm [800 gpm]) into the RPV at operating 

pressure and is located outside of the reactor building. The system is designed to be capable of 

providing sufficient core cooling in the unlikely event that all normal and emergency core cooling 

systems are unavailable. The power supply for the AFI system is non-safety-related and is 

independent and physically separated from the emergency power supplies such that a 

simultaneous loss due to a beyond design basis event is unlikely. 

The following I&C signals are provided in the AFI pump house: 

• RPV water level 

• RPV pressure 

• Wetwell wide range (WR) pressure 

• Suppression pool water level 

• AFI pump flow and discharge pressure 

• AFI dedicated water storage tank water level 
7-2

7.7.3 Regulatory Basis 

The regulatory basis and acceptance criteria for reviewing the instrumentation and controls not 

required for safety are in Section 7.7 of NUREG–0800. 

In particular, the following acceptance criteria are applicable to this review: 

a. 10 CFR 50.55a(h), “Protection and safety systems,” requires compliance with IEEE 

Std 603-1991, "IEEE Standard Criteria for Safety Systems for Nuclear Power Generating 

Stations," and the correction sheet dated January 30, 1995. For control systems isolated 

from safety systems, the applicable requirements of 10 CFR 50.55a(h) are defined in 

IEEE Std 603-1991 Clause 5.6.3, "Independence Between Safety Systems and Other 

Systems," and IEEE Std 603-1991 Clause 6.3, "Interaction Between the Sense and 

Command Features and Other Systems." 

b. GDC 24, “Separation of protection and control systems,” states that the protection 

system shall be separated from control systems to the extent that failure of any single 

control system component or channel, or failure or removal from service of any single 

protection system component or channel which is common to the control and protection 

systems leaves intact a system satisfying all reliability, redundancy, and independence 

requirements of the protection system. Interconnection of the protection and control 

systems shall be limited so as to ensure that safety is not significantly impaired. 

c. 10 CFR 52.47(b)(1) requires that a design certification (DC) application contain the 

proposed inspections, tests, analyses, and acceptance criteria (ITAAC) that are 

necessary and sufficient to provide reasonable assurance that, if the inspections, tests, 

and analyses are performed and the acceptance criteria met, a plant that incorporates 

the design certification has been constructed and will operate in accordance with the 

design certification, the provisions of the Atomic Energy Act, and the NRC's regulations. 

7.7.4 Technical Evaluation 

The purpose of the staff’s evaluation is to assess the impact of the AFI system’s I&C on the 

plant’s existing control and protection systems. The AFI system’s I&C is independent and 

separate from the normal plant control and protection systems. However, the field sensors 

associated with the AFI system share process connections with safety-related plant I&C 

sensors. The staff’s evaluation is to verify that the normal plant safety and non-safety-related 

I&C systems are not adversely impacted by the interfacing AFI system’s I&C. 

10 CFR 50.55a(h), which requires compliance with IEEE Std 603-1991, and GDC 24 require a 

non-safety-related system should be designed so as to not prevent a safety system from 

performing its intended function and to prevent the effects of a single random failure. The staff 

asked the applicant to demonstrate that the I&C for the AFI system does not adversely impact 

the plant’s existing safety-related and important-to-safety I&C systems. In response to 

RAI 07.07-1, dated March 3, 2010 (U7-C-STP-NRC-100056, ML100640162), the applicant 

states that there are no automatic controls or functions associated with the AFI system, and it’s 

I&C are hard-wired and manually initiated. In addition, the applicant states that the AFI 

system’s I&C is only used for indication purposes, and the water level and pressure 

instrumentation in the AFI pump room use a separate set of transmitters and a separate power 

supply independent of the existing I&C. As part of the response to RAI 07.07-1, the applicant 

also provides changes to the AIA DCD amendment application to clarify the safety classification 
7-3

of the I&C components in the AFI system, which the staff found acceptable. The applicant’s 

response to RAI 07.07-1, which references RAI 03.02.02-1, states that the safety classification 

of the AFI I&C sensors and piping is the same as that of the existing interfacing I&C system. 

The applicant’s response to RAI 03.02.02-1 provides a markup of DCD Tier 2, Table 3.2-1, 

“Classification Summary,” which now includes the AFI system’s I&C and associated piping with 

Seismic Category I classification. Therefore, the staff found that the Seismic Category I AFI 

transmitters that share the safety-related process connections will not adversely impact the 

plant’s existing safety functions. The staff found that the DCD amendment complies with the 

acceptance criteria for 10 CFR 50.55a(h) and GDC 24. The staff reviewed the revised DCD 

AIA Amendment and confirmed that the proposed changes are incorporated. Therefore, 

RAI 07.07-1 and RAI 03.02.02-1 are resolved and closed. 

The staff asked the applicant to provide an ITAAC in the DCD amendment to demonstrate that 

the I&C included in the AFI system will not adversely affect and is adequately isolated from the 

plant safety systems. The applicant’s response to RAI 07.07-3 dated April 8, 2010 

(U7-C-STP-NRC-100049, ML101040254), states that the safety classification of the I&C and 

piping for the AFI system is the same as that for the existing plant I&C to which it is connected. 

The applicant’s response to RAI 03.02.02-1 provides a markup of DCD Tier 2, Table 3.2-1, 

“Classification Summary,” which includes the AFI-related I&C classified as Seismic Category I. 

The applicant’s response to RAI 07.07-3 and RAI 14.02-1 dated April 8, 2010 

(U7-C-STP-NRC-100049, ML101040254), provides a markup of the ABWR DCD Tier 1 that 

includes an ITAAC for the AFI system. ITAAC Table 2.11.24, Item 12 includes a design 

commitment to verify that an AFI instrumentation device that is physically attached to 

instrumentation piping satisfies the same requirements (safety class, quality group, and seismic 

category) as the instrumentation piping to which it is attached. Also in this response, the 

applicant references an existing ITAAC (Table 3.3, Item 1) verifying that the AFI instrumentation 

piping meets the quality requirements identified in Tier 2, Table 3.2-1. The staff found that 

these ITAAC are adequate to verify that the AFI system’s I&C will not have an adverse impact 

and is adequately isolated from the plant’s existing control and protection systems, and 

therefore are in compliance with 10 CFR 52.47(b)(1). The staff reviewed the revised DCD AIA 

Amendment and confirmed that the proposed changes are incorporated. Therefore, 

RAI 03.02.02-1, RAI 07.07-3 and RAI 14.02-1 are resolved and closed. 

7.7.5 Conclusion 

NRC staff reviewed the AFI system’s I&C of the STPNOC's application to amend the Design 

Certification Rule for the ABWR. The staff performed an evaluation to confirm that the AFI 

system’s I&C interface with the plant’s safety and control systems conforms to the applicable 

regulations and does not adversely impact the plant’s existing safety and control functions. The 

staff used applicable criteria discussed in Section 7.7.3 of this SER to evaluate the AFI system’s 

I&C. The staff found that the AFI system’s I&C described in the DCD amendment satisfies 

the applicable acceptance criteria of SRP Section 7.7, which address 10 CFR 50.55a(h), 

10 CFR 52.47(b)(1), and GDC 24, and does not adversely impact the plant’s existing safety and 

control systems. The staff determined that there would be no adverse impacts from complying 

with the requirements for consideration of aircraft impacts on conclusions reached by the NRC 

in its review of the original ABWR design certification (NUREG-1503).

9-1

9 AUXILIARY SYSTEMS 

9.5.1 Fire Protection Program 

9.5.1.1 Introduction

This section describes the NRC staff’s evaluation of the changes to the credited fire protection 

program within the certified ABWR design. This section of the safety evaluation discusses the 

staff’s evaluation of fire protection with the proposed changes and the technical basis for the 

staff’s acceptance of these proposed changes. This section also discusses the staff’s 

determination of whether the changes adversely impact conclusions reached by the NRC in its 

review of the original ABWR design certification. 

9.5.1.2 Summary of Application 

The applicant has submitted a design certification amendment incorporating several fire 

protection program changes. These fire protection-related design changes are reported in the 

amended ABWR DCD Tier 2, Revision 3, Appendix 19S, with further details described in 

Section 9.5.1 and Appendix 9A that include Figures 9A-1 through 9A-11. These changes 

include the following: 

• a non-fire-rated floor becoming a rated 3-hour fire barrier 

• addition of fire-rated floor plugs 

• several regular fire doors becoming capable of withstanding a 5 psid 

• all fire barriers, including associated penetration seals, within the reactor building 

capable of withstanding 5 psid 

• several fire areas split into two or more fire areas 

9.5.1.3 Regulatory Basis

The regulatory basis for reviewing the fire protection program is in Section 9.5.1 of 

NUREG-0800, Revision 5. The relevant requirements are: 10 CFR 50.48 and 10 CFR Part 50, 

Appendix A, GDC 3, and guidance of RG 1.189, Revision 1. 

9.5.1.4 Technical Evaluation

NRC staff reviewed the description of certified design changes relating to fire protection. The 

proposed changes from a non-rated barrier to a fire rated barrier or the splitting of a fire area 

into smaller fire areas increases the compartmentation, reduces the spread of fire, and 

increases protection of safe shutdown equipment. Changing a regular 3-hour fire barrier to a 

barrier with the same fire rating plus an increased pressure capability reduces the spread of fire 

and increases the protection of safe shutdown equipment. The applicant’s changes do not 

invalidate or reduce the fire protection capabilities of the certified design, but rather increase the 

protection worth.
9-2

9.5.1.5 Conclusion

NRC staff reviewed the fire protection system of the STPNOC's application to amend the Design 

Certification Rule for the ABWR. NRC staff’s review concluded that the applicant has 

adequately addressed the requirements of 10 CFR 50.48 and 10 CFR Part 50, Appendix A, 

GDC 3. The staff determined that there would be no adverse impacts from the amended design 

on conclusions reached by the NRC in its review of the original ABWR design certification 

(NUREG-1503). Therefore, staff found these changes acceptable. 

9.5.14 Alternate Feedwater Injection System 

9.5.14.1 Introduction

As part of the application, STPNOC proposes to add an AFI system. The AFI system is a 

reactor coolant makeup system capable of providing sufficient core cooling in the unlikely event 

that all normal and emergency core cooling systems are unavailable. The system has a 

dedicated water source and injection is provided through the non-safety-related portion of the 

CUW tie-in lines to the feedwater system. The AFI system capacity and discharge pressure (at 

rated pressure) are comparable to the same parameters in the high-pressure core flooder 

(HPCF) system. This section of the safety evaluation discusses the staff’s evaluation of the 

proposed additions and the technical basis for the staff’s acceptance of these proposed 

changes. Also, this section discusses the staff’s determination of whether the changes 

adversely impact conclusions reached by the NRC in its review of the original ABWR design 

certification. 

9.5.14.2 Summary of Application

DCD Revision 3, Tier 2, Section 9.5.14, “Alternate Feedwater Injection System,” describes the 

AFI system. The system is capable of injecting into the RPV at operating pressure and 

providing sufficient core cooling in the event that normal and emergency cooling systems are 

not available. 

A schematic of the system is depicted in DCD Tier 2, Figure 9.5-6. The system is in the AFI 

pump house. The AFI system takes suction from a water source located near the AFI pump 

house and discharges through three normally closed motor-operated valves into the feedwater 

system through the non-safety-related portion of the CUW tie-in lines to the feedwater system. 

DCD Revision 3, Tier 2, Table 6.2-9, “Secondary Containment Penetration List,” identifies two 

new penetrations related to AFI system in the secondary containment. 

The AFI system and its power supplies are non-safety grade, except for portions of the I&C (as 

discussed in Section 7.7 of this SER). The power supply for the pump and motor-operated 

valves are non-safety-related, and are physically separated from the emergency power 

supplies. 

The system can be operated from the AFI pump house. The applicant states that this will 

ensure that the injection can be initiated within 30 minutes after the loss of normal makeup 

systems to provide sufficient cooling. In addition, the operator is provided with capability to 

control flow from the AFI pump house by throttling a motor-operated valve located in the pump 

room. 
9-3

In the initial DCD Amendment Table 1.9-1, “Summary of ABWR Standard Plant COL License 

Information,” the applicant identified the following COL license information items: 

ITEM NO SUBJECT

9.39 Power Supply for Alternate Feedwater Injection Equipment

9.40 Test and Surveillance Intervals for Alternate Feedwater Injection Equipment,

In the DCD amendment dated May 12, 2010 (ML101340548) the applicant deleted the COL 

license information items listed above. 

ITAAC: The discussion of ITAAC is in Section 19S.4.5 of this SER. 

Technical Specifications: There are no technical specification (TS) requirements associated 

with the AFI System. 

9.5.14.3 Regulatory Basis 

The regulatory basis and acceptance criteria for reviewing the AFI system are summarized 

below: 

1. GDC 2, “Design bases for protection against natural phenomena,” as it relates to 

safety-related SSCs that must be protected from, or capable of withstanding, the effects of 

such natural phenomena as earthquakes, tornadoes, hurricanes, and floods as described in 

DCD Chapters 2 and 3. 

2. GDC 4, “Environmental and dynamic effects design bases,” as it relates to safety-related 

SSCs that must be protected from, or capable of withstanding, the effects of externally 

generated missiles, internally generated missiles, pipe whip and jet impingement forces 

associated with pipe breaks and dynamic effects associated with possible fluid flow 

instabilities (e.g., water hammers). 

3. GDC 60, “Control of releases of radioactive materials to the environment,” as it relates to the 

ability of the AFI system design to control releases of radioactive materials to the 

environment.

4. GDC 16 “Containment design,” as it relates to reactor containment and associated systems 

establishing an essentially leak-tight barrier against the uncontrolled release of radioactivity 

to the environment. 

9.5.14.4 Technical Evaluation 

NRC staff reviewed the STPNOC’s application to amend the ABWR Design Certification Rule 

of the ABWR DCD. The staff’s review of the proposed COL License Information Items 9.39 

and 9.40 in the initial DCD amendment determined that these items should be part of the 

ITAAC. The applicant’s revised DCD amendment withdrew these COL license information 

items and included additional ITAAC. These changes are consistent with the staff’s 

recommendation, and therefore acceptable. 

The staff’s acceptance of the AFI system is based on the compliance of the system design with 

the requirements of GDC 2, 4, 16, and 60. 
9-4

The results of the staff’s review are discussed below. 

GDC 2, “Design bases for protection against natural phenomena” 

The staff reviewed the AFI system for compliance with the requirements of GDC 2 with respect 

to the design for protection against the effect of natural phenomena such as earthquakes, 

tornados, hurricanes, and floods. Compliance with the requirements of GDC 2 is based on 

adherence to Position C.1 of RG 1.29, “Seismic Design Classification,” for the safety-related 

portions of the system and Position C.2 for non-safety-related portions of the system. 

The applicant states in Subsection 9.5.14.1 of the amendment that the AFI system 

discharge piping is routed underground or is otherwise protected from physical impact. 

Subsection 9.5.14.2 of the amendment states that the piping and components interfacing with 

the CUW system are the same quality as that of the system up to and including the second 

check valve. 

The AFI system will be located in AFI pump house and its water source will be located at least 

300 feet from the reactor building, control building and turbine building. The AFI system is nonsafety-related, and the AFI pump house contains no safety-related SSCs. Therefore, any failure 

of the system housed in the AFI pump house will not result in the failure of any safety-related 

SSCs and will not adversely affect the safety-related function of the ABWR design. 

The AFI water source, which is located outside of the AFI pump house, will contain a minimum 

of 1,136,000 liters (300,000 gallons) of water for the AFI system to use. There is no information 

included in the amendment on what impact the failure of this tank would have on the plant 

flooding analysis. Because the water source supplying the AFI system contains a large 

inventory and is non-seismic, the staff issued RAI 09.02.04-2 requesting the applicant to discuss 

the impact that a failure of the water source would have on safety-related equipment. 

The applicant’s response to RAI 09.02.04-2 dated February 25, 2010 (U7-C-STP-NRC-100040, 

ML100600410), addresses potential flooding from the AFI water source by stating that the water 

source for the AFI is required to be located at least 90 meters (300 feet) from the reactor 

building, and the separation will minimize the possibility that flooding associated with the failure 

of the 1,136,000 liters (300,000 gallon) water source will have on SSCs located in those 

buildings. The applicant adds that in ABWR DCD Subsection 3.4.1.1.1, the flood protection 

measures in the ABWR design guard against flooding from outside storage tanks that may 

rupture, which include the condensate storage tank (CST), which has a capacity twice that of 

the AFI water source. The applicant concludes that any flooding from the AFI water source is 

bounded by the existing flood analysis. 

The staff reviewed the flood analysis in DCD Subsection 3.4.1.1 and agreed that external 

flooding from an AFI water source failure is bounded by the external flooding that could result 

from the possible failure of the CST that was analyzed in the flood analysis. The staff’s concern 

in RAI 09.02.04-2 regarding the AFI water source failure is therefore resolved. 

The AFI system discharges through three normally closed MOVs into the feedwater system 

through the non-safety-related portion of the CUW tie-in lines, which are located in the reactor 

building main steam tunnel. Because, the break of the AFI line in the steam tunnel could also 

have flooding implications, RAI 09.02.04-2 also requests the applicant to discuss how the AFI 

line break is addressed in the appropriate flood analysis. 
9-5

The applicant’s response to RAI 09.02.04-2 dated February 25, 2010 (ML100600410), states 

that an AFI line break in the main steam tunnel area is bounded by the feedwater high energy 

line break, which is analyzed in DCD Tier 2, Subsection 3.4.1.1.2. The applicant further 

indicates that a break in the main steam tunnel area will be contained in the Seismic Category I 

structure and will not flood any safety-related equipment in the reactor building and control 

building. 

The staff reviewed the applicant’s response and the flooding analysis information in DCD Tier 2, 

Subsection 3.4.1.1.2 and found that the response adequately addresses the staff’s concern, 

because the bounding break in the main steam tunnel is the feedwater line break and if flooding 

occurs in the main steam tunnel, the water volume is kept inside the tunnel until operators are 

ready to pump it to the radwaste system. Therefore, no safety-related equipment will be 

affected. The applicant provides a markup of DCD Section 9.5.14.2 “Safety Evaluation” to 

incorporate the above changes into the DCD. The staff confirmed that the revised ABWR DCD 

AIA Amendment, Subsection 9.5.14.2 is revised as committed in the RAI response. The staff 

considers RAI 09.02-04-2 resolved and closed. 

Based on the above review, the staff concluded that the AFI system meets the requirements of 

GDC 2 as it relates to protecting the system against the effects of natural phenomena. 

GDC 4, “Environmental and dynamic effects design bases” 

The staff reviewed the AFI system for compliance with the requirements of GDC 4, 

“Environmental and dynamic effects design bases,” as they relate to the dynamic effects 

associated with possible fluid flow instabilities, including induced water hammer and the effects 

of pipe breaks. Compliance with the requirements of GDC 4 is based on identification of the 

essential portions of the system that are protected from dynamic effects, including internally and 

externally generated missiles, pipe whip and jet impingement due to high and moderate energy 

missiles, and meeting the guidance in Branch Technical Position (BTP) 10-2, “Design 

Guidelines for Avoiding Water Hammers in Steam Generators.” 

The AFI system is a non-safety-related system. The system is being reviewed against GDC 4 

to ensure that safety-related SSCs are not affected due to the AFI system’s operation or its 

proximity to safety-related SSC’s. The applicant states in Subsection 9.5.14.1 of the DCD 

amendment application that the AFI system discharge piping is routed underground or is 

otherwise protected from physical impact. Subsection 9.5.14.2 of the amendment states that 

the piping and components that interface with the CUW system are the same quality as those of 

the system up to and including the second check valve. 

The application does not address what impact if any the failure of an AFI line could have on 

safety-related equipment in the reactor building or main steam tunnel, or what design features 

or operational measures will be used to minimize or preclude water hammer events due to the 

operation of the AFI system. The staff requested the applicant to provide information to address 

the AFI line failure (RAI 09.02.04-4) and water hammer (RAI 09.02.04-3). 

The applicant’s response to RAI 09.02.04-4 dated February 25, 2010 (ML100600410), states 

that regarding the impact of a postulated AFI system break on safety-related systems, the AFI is 

only required for beyond-design-basis events and is powered off and unpressurized for all 

normal and design-basis events. The applicant also states that should a pipe break in the AFI 

system occur, the dynamic effects of the break on the plant’s safety systems are bounded by 

the main steam and feedwater line breaks because the AFI lines are at much lower temperature 
9-6

and pressure and have smaller line sizes. The staff found the applicant’s response acceptable, 

since it confirms that the system is normally isolated and unpressurized, and that other breaks 

(main steam and feedwater) are bounding in terms of the resulting dynamic effects. Therefore, 

RAI 09.02.04-4 is resolved and closed. 

The applicant’s response to RAI 09.02.04-3 dated February 25, 2010 (ML100600410), states 

that the AFI design includes a “keep–full” line to maintain the system full of water to preclude 

water hammer upon system initiation. The applicant also states that existing vent valves in the 

feedwater pipe can be used to fill and vent the AFI system to ensure that the piping is 

maintained full of water. The applicant states the AFI system will be added to the ABWR DCD 

Section 13.5 list of systems for which operating procedures will need to be written. The 

procedures will include provisions to throttle the flow at the start of the AFI pump to minimize the 

potential for water hammer. 

In this response, the applicant also provides a markup of DCD Subsection 9.5.14.1, “System 

Description,” and DCD Subsection 13.5.3.4, “Procedures included in Scope of Plan,” to 

incorporate these changes into the revised DCD. The staff found the applicant’s response to 

RAI 09.02.04-3 acceptable, because the applicant adequately addresses water hammer 

concerns with system design features (“keep-full” line) and through the use of procedural 

precautions used in the industry to reduce or minimize water hammer. The staff confirmed that 

the revised ABWR DCD AIA Amendment, Subsections 9.5.14.1 and 13.5.3.4 are revised as 

committed in the RAI response. The staff considers RAI 09.02.04-3 resolved and closed. 

Based on the above review, the staff concluded that the AFI system meets the requirements of 

GDC 4 as they relate to the dynamic effects associated with possible fluid instabilities, including 

induced water hammer and the effects of pipe breaks. 

GDC 60, “Control of releases of radioactive materials to the environment,”

NRC staff reviewed the design of the AFI system for compliance with the requirements of 

GDC 60 with respect to control of releases of radioactive materials. GDC 60 requires that a 

means be provided to control the release of radioactive materials in liquid effluents. 

Information on how the control of release of radioactive materials to environment is 

accomplished for the AFI system was not included in the amendment. Therefore, the staff 

issued RAI 09.02.04-1 requesting the applicant to address how the AFI system complies with 

GDC 60. 

The applicant’s response to RAI 09.02.04-1 dated April 19, 2010 (U7-C-STP-NRC-100085, 

ML101120085), states that the AFI contains two check valves located in the reactor building 

main steam tunnel, and three normally-closed MOVs. The applicant adds that in the event that 

fluid from the CUW system should leak past the two check valves and the first MOV, a leakoff 

line is included in the AFI design that directs any leakage back to the reactor building’s lowconductivity sump, and the existing leak detection and radiation monitoring can be used to 

monitor the leakage. The applicant provides a markup of DCD Section 9.5.14.2 “Safety 

Evaluation” to incorporate the above changes into the DCD. The staff confirmed that the 

revised ABWR DCD AIA Amendment, Subsection 9.5.14.2 is revised as committed in the RAI 

response. The staff considers RAI 09.02.04-1 resolved and closed. 

The staff found the information provided by the applicant in the response to RAI 09.02.04-1 

acceptable, because the applicant design provides barriers against radiation leakage from 
9-7

radioactive systems into the non radioactive AFI system. The design also provides for routing of 

AFI system leakage to the reactor building low conductivity sumps, and existing leak detection 

and radiation monitoring can be used to monitor the leakage. 

Based on the information in the applicant’s response to RAI 09.02.04-1 and the staff’s review of 

Figure 9.5-6 of the amendment, the staff found that the AFI design meets the requirements of 

GDC 60, as they relate to the AFI having design provisions to control the release of radioactive 

materials into the environment. 

GDC 16 “Containment design” 

In addition, the staff reviewed the design for impact on the integrity of the reactor building which 

forms the secondary containment under GDC 16 as it relates to establishing an essentially 

leak-tight barrier against the uncontrolled release of radioactivity to the environment. The 

additional penetrations identified in Table 6.2-9 are fluid piping penetrations that are designed 

under containment isolation provisions. This change does not alter the staff conclusion as 

documented in NUREG-1503. The staff found the addition of two penetrations acceptable. 

Technical Specifications: The DCD amendment application does not include a TS section for 

the AFI system. The staff reviewed the AFI system against 10CFR 50.36, “Technical 

specifications,” and agreed that no additional TS are required for this system. Therefore, the 

staff found this aspect of the DCD acceptable. 

9.5.14.5 Conclusion 

NRC staff reviewed the AFI system as part of the STPNOC's application to amend the Design 

Certification Rule for the ABWR. The staff’s review of the information in the application found 

that it provides adequate information supporting the AFI system functional design. The staff 

found the amended design acceptable because it meets the appropriate regulatory 

requirements, including GDC 2, protection from natural phenomena; GDC 4, protection against 

missiles and effects of pipe break; and GDC 60, control of releases of radioactive materials into 

the environment. NRC staff also reviewed the two new penetrations identified in Table 6.2-9. 

The staff found the change acceptable because it meets the regulatory requirements of 

GDC 16. The staff determined that there would be no adverse impacts from complying with the 

requirements for consideration of aircraft impacts on conclusions reached by the NRC in its 

review of the original ABWR design certification (NUREG-1503).

13-1

13 CONDUCT OF OPERATIONS 

13.5.3.4 Procedures Included in Scope of Plant 

13.5.3.4.1 Introduction 

As part of the application, STPNOC proposes to include the AFI system within the scope of the 

plant operating procedures development plan so that appropriate procedures will be developed 

for the AFI system. This section of the SER discusses the staff’s evaluation of the proposed 

changes and the technical basis for the staff’s acceptance of these proposed changes. 

13.5.3.4.2 Summary of Application 

In the initial submittal dated June 30, 2009 (ML092040048), the applicant proposed to include 

the systems described in Section A.3 of ANSI/ANS-3.2, within the scope of the plant operating 

procedures development plan. 

To address the aircraft impact assessment required by 10 CFR 50.150, the applicant, in a letter 

dated June 17, 2010 (U7-C-STP-NRC-100139, ML101720306), proposes to include the AFI 

system within the scope of the plant operating procedures development plan. 

13.5.3.4.3 Regulatory Basis 

The regulatory basis is established in 10 CFR Part 52, and Appendix B to 10 CFR Part 50; the 

guidance of RG’s 1.8, 1.33, 1.70, and 1.206; and the guidance of Section 13.5.2.1 of 

NUREG-0800, Revision 2, and ANSI/ANS-3.2. 

13.5.3.4.4 Technical Evaluation

NRC staff reviewed the information related to the AIA application in DCD Revision 3. 

Section A.3 of ANSI/ANS-3.2 identifies systems to which the guidance of ANSI/ANS-3.2 is 

applicable. The applicant is adding one system in response to the aircraft impact assessment, 

which is required by 10 CFR 50.150. This new system is properly identified in accordance with 

ANSI/ANS 3.2, and Section A.3 is the appropriate place to include the new system. Also, this is 

acceptable to comply with 10 CFR 50.120, “Training and qualification of nuclear power plant 

personnel.” 

13.5.3.4.5 Conclusion 

NRC staff reviewed the inclusion of the AFI system within the scope of the plant operating 

procedures development plan. The staff’s review found the applicant has adequately addressed 

the applicable requirements. In addition, the staff determined that there would be no adverse 

impacts from complying with the requirements for consideration of aircraft impacts on 

conclusions reached by the NRC in its review of the original ABWR design certification 

(NUREG-1503).

14-1

14 INITIAL TEST PROGRAM 

14.2.1 Initial Test Program (ITP) 

14.2.1.1 Introduction 

As part of the application, STPNOC proposes changes the Initial Test Program (ITP) 

preoperational tests. This section of the ABWR DCD addresses the series of tests categorized 

as construction, preoperational, or initial startup tests. The application to amend the Design 

Certification Rule for the ABWR changes the ITP preoperational tests to meet the requirements 

in 10 CFR 52.79(a)(28), and 10 CFR 50.150 as they relate to STPNOC’s request to amend the 

ABWR standard design certification to comply with the AIA rule. This section of the safety 

evaluation discusses the staff’s evaluation of the ITP of the proposed additions and the 

technical basis for the staff’s acceptance of these proposed changes. 

14.2.1.2 Summary of Application 

NRC staff identified the following changes in DCD Section 14.2, “Initial Plant Test Program,” to 

address testing requirements for AFI equipment. 

Alternate Feedwater Injection System Preoperational Test

In Subsection 14.2.12.1.78, “AFI Preoperational Test,” the ABWR DCD AIA Amendment 

application incorporates the “Purpose, Prerequisites, General Test Methods and Acceptance 

Criteria” for this test. 

The purpose of the AFI system preoperational test is to include testing of equipment, pumps, 

valves, and I&C. The prerequisites for conducting this preoperational test include completion of 

construction tests, The Startup Control Group (SCG) review and approval of the test procedure 

and of the test initiation. This includes the availability of an AFI pump suction water source to 

the reactor vessel through feed water lines A and B to receive AFI injection flow, and verification 

of the availability of an appropriate electrical power source to support this test. 

The General Test Methods and Acceptance Criteria include performance of tests on individual 

components in the AFI system. The tests are designated to demonstrate that the AFI system 

operates properly through appropriate AFI system design specifications using the following 

12 test methods and acceptance criteria: 

• Proper operation of I&C in the AFI system. 

• Verification of various component alarms. 

• Proper operation of all MOVs, including opening and closing the MOVs with the operating 

switch, MOV status indication, and travel timing, if applicable. 

• Proper operation of AFI pumps and motors during continuous run tests. 

• Acceptable pump net positive suction head (NPSH) under the most limiting design flow 

conditions. 
14-2

• Verification that the AFI System can be operated normally at each mode and will satisfy the 

NPSH requirement by combining all components, piping, and instruments constituting this 

system. 

• Proper AFI pump motor start sequence and actuation of protective devices. 

• Proper operation of interlocks including operation of all components subject to interlocking. 

• Proper operation of permissive, prohibit, and bypass functions. 

• Proper system operation while powered from primary and alternate sources, including 

transfers, and in degraded modes for which the system is expected to remain operational. 

• Acceptable pump/motor vibration levels and system piping movements during both transient 

and steady-state operations. This test can be performed in conjunction with the expansion, 

vibration, and dynamic effects preoperational test (Subsection 14.2.12.1.51). 

• Proper operation of the pump discharge line keep-fill system and its ability to prevent 

damaging water hammer during system transients. 

14.2.1.3 Regulatory Basis

The regulatory basis and acceptance criteria for reviewing ITP are in Section 14.2 of 

NUREG-0800. In particular, the applicable regulatory requirements and guidance for the ITP 

portion of the ABWR DCD Amendment Include 10 CFR 52.79(a)(28), as it relates to 

preoperational testing and initial operations; and the guidance in RG 1.68, “Initial Test 

Programs For Light Water-Cooled Nuclear Power Plant,” dated March 2007 and RG 1.206, 

“Combined License Applications for Nuclear Power Plants,” Section C.I.14, “Verification 

Programs,” dated June 2007. 

14.2.1.4 Technical Evaluation

In accordance with 10 CFR 52.79(a)(28), NUREG–0800 Section14.2, and the regulatory 

positions in RG 1.68 and RG 1.206 Section C.I.14, the NRC staff reviewed the DCD 

amendment for these proposed changes to preoperational tests in Tier 2, DCD Section 14.2, to 

verify that they will satisfy the ITP. 

Based on the regulations in 10 CFR 52.79(a)(28), the guidance in NUREG–0800. Section 14.2, 

and the regulatory positions in RG 1.68 and RG 1.206, the staff evaluated the AFI 

preoperational test acceptance criteria. The staff determined that, the purpose, prerequisites, 

and 12 general test methods and acceptance criteria in Tier 2, DCD Subsection 14.2.12.1.78 

met 10 CFR 52.79(a)(28), RG 1.68 and RG 1.206, Section C.1.14 since they adequately test 

feedwater design features needed to mitigate the AIA beyond design basis event; therefore; 

they are acceptable. 

14.2.1.5 Conclusion

NRC staff reviewed the ITP of the STPNOC's application to amend the Design Certification Rule 

for the ABWR. The staff’s review concluded that the addition of Subsection 14.2.12.1.78, “AFI 

System Preoperational Test,” met the requirements in 10 CFR 50.150, 10 CFR 52.79(a)(28), 
14-3

and conforms with the guidance in NUREG–0800 Section 14.2 and the regulatory positions in 

RG 1.68 and RG 1.206 Section C.I.14. Therefore, the test is acceptable. 

14.3 Tier 1 Selection Criteria and Processes 

14.3.1 Introduction

As part of the application, STPNOC proposes to add ITAAC for the AFI system and to specify 

the protection afforded by the reactor building in DCD Tier 1. The Tier 1 information in the 

ABWR DCD consists of an introductory section; design descriptions and corresponding ITAAC 

for the systems of the design, design material applicable to multiple systems of the design, 

interface requirements, and site parameters for the ABWR design. Furthermore, the purpose of 

the ITAAC, which are part of the Tier 1 information, is to verify that a facility that references the 

design certification has been constructed and will operate in accordance with the design 

certification and the applicable requirements. 

14.3.2 Summary of Application 

On May 12, 2010, STPNOC submitted a revised application to amend the Design Certification 

Rule for the U.S ABWR. In this submittal, the applicant proposes changes to Section 2.11.24, 

“Alternate Feedwater Injection System,” and Section 2.15.10, “Reactor Building,” of the certified 

DCD Tier 1 Information. 

14.3.3 Regulatory Basis 

The regulatory basis and acceptance criteria for reviewing the ITAAC information are in 

Section 14.3 of NUREG–0800. In particular, the applicable regulatory requirements and 

guidance for the ITAAC portion of the ABWR DCD Amendment Include: 10 CFR 50.150, and 

10 CFR 52.47(b)(1), as they relate to ITAAC; and the guidance in RG 1.206, “Combined 

License Applications for Nuclear Power Plants,” Section C.I.14, “Verification Programs,” dated 

June 2007. 

14.3.4 Technical Evaluation 

In the applicant’s initial DCD amendment, the staff found that the applicant did not propose 

an ITAAC for the AFI system. The staff issued RAI 14.2-01 requesting the applicant to provide 

an ITAAC for the AFI system. In response to this RAI, the applicant revised DCD Tier 1, 

Section 2.11.24 and Table 2.11.24 to meet the ITAAC requirements in 10 CFR 52.47(b)(1). 

These ITAAC are evaluated in Section 19S.4.5 of this SER. 

In Tier 1, DCD Section 2.15.10, the applicant added a statement emphasizing that the specific 

areas of the reactor building are analyzed for the effects of postulated impact of an aircraft as 

required by 10 CFR 50.150. The staff reviewed the applicant’s statement and concluded that 

1) it was consistent with other Tier 1 information provided for the reactor building, 2) it was 

consistent with associated Tier 2 information, and 3) it adequately described the performance 

characteristics of the reactor building in relation to aircraft impacts. Therefore, the staff found the 

additional Tier 1 information acceptable. 
14-4

14.3.5 Conclusion 

The staff reviewed and concluded that the applicant proposed ITAAC for the AFI system are 

acceptable to meet the requirements in 10 CFR 50.150, and 10 CFR 52.47(b)(1) and conforms 

with the guidance in NUREG–0800, Section 14.3. For additional details, see section 19S.4.5 of 

this SER. 

The staff reviewed and concluded that the applicant’s proposed change in the DCD Tier 1, 

Section 2.15.for the reactor building is an appropriate Tier 1 description and is acceptable. 
17-1

17 QUALITY ASSURANCE 

17.1.19 Quality Assurance During the Design Phase 

17.1.19.1 Introduction

As part of the application, STPNOC proposes revising Section 17.0, “Introduction,” of the 

ABWR DCD and adding Section 17.1.19, “Aircraft Impact Assessment,” to Section 17.1, “Quality 

Assurance During the Design Phase.” STPNOC also submitted an application on 

September 20, 2007, for a COL for two ABWRs designated as STP Units 3 and 4. The 

proposed Section 17.1.19 of the amended DCD references the STP Units 3 and 4 QA Program 

Description (QAPD). However, this SER does not apply to the STP Units 3 and 4 COL 

application. 

This section of the safety evaluation discusses the staff’s evaluation of the QA program during 

the design phase of the proposed additions and the technical basis for the staff’s acceptance of 

these proposed changes. Also this section discusses the staff’s determination of whether the 

changes adversely impact conclusions reached by the NRC in its review of the original ABWR 

design certification. 

17.1.19.2 Summary of Application 

The proposed Section 17.1.19, “Aircraft Impact Assessment,” of the amended DCD Tier 2, 

Revision 3 addresses the establishment and implementation of a QA Program applicable to the 

proposed amendment to the ABWR design certification. In Section 17.1.19, the applicant 

references the STP Units 3 and 4 QAPD, Revision 4, dated June 7, 2010. The STP Units 3 

and 4 QAPD, Revision 4, is included in the revised ABWR DCD AIA amendment dated 

June 17, 2010, (ML101720306). 

17.1.19.3 Regulatory Basis 

The regulatory basis for accepting the resolution to STP DCD Tier 2, Revision 3, 

Section 17.1.19 is satisfied based on the Commission’s regulatory requirements related to QA 

Programs, which are set forth in 10 CFR 52.47(a)(19) and Appendix B, “Quality Assurance 

Criteria for Nuclear Power Plants and Fuel Reprocessing Plants,” to 10 CFR Part 50. 

10 CFR 52.54(a)(8) requires that the applicant implement the quality assurance program 

described or referenced in the safety analysis report. 

10 CFR 52.47(a)(19) requires, in part, that a DC application contains a description of the QAP 

applied to the design of the SSCs of the facility. 10 CFR 52.47(a)(19) further requires that the 

description of the QAP for a nuclear power plant include a discussion of how the applicable 

requirements of Appendix B were satisfied. 

10 CFR Part 50, Appendix B establishes QA requirements for the design, fabrication, 

construction, and testing of safety-related SSCs of the facility. The pertinent requirements of 

Appendix B apply to all activities affecting the safety-related functions of those SSCs and 

include designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, 

installing, inspecting, testing, operating, maintaining, repairing, refueling, and modifying SSCs. 
17-2

17.1.19.4 Technical Evaluation 

RAI 01-2 and the applicant’s response to this RAI are described in Section 1.4 of this SER. 

STPNOC’s supplemental response to RAI 01-2 dated April 8, 2010 (U7-C-STP-NRC100080, 

ML101040345), describes the applicant’s QA approach to the aircraft impact assessment and to 

the design changes resulting from that assessment. The applicant also proposes revising 

ABWR DCD Section 17.0, “Introduction,” and adding Section 17.1.19, “Aircraft Impact 

Assessment,” to Section 17.1, “Quality Assurance During the Design Phase.” These proposed 

changes are incorporated into the ABWR DCD AIA amendment dated June 17, 2010, 

(ML101720306). 

NRC staff reviewed (a) the proposed DCD Tier 2 Revisions provided in the supplemental 

response to RAI 01-2; (b) STPNOC’s application to amend the Design Certification Rule for the 

U.S. ABWR; and (c) STP Units 3 and 4 QAPD, Revision 3. The staff’s review is based on 

10 CFR Part 50, Appendix B and applicable provisions of SRP Section 17.5. SRP Section 17.5 

provides an outline of a QA Program acceptable to the staff for DC applicants. The staff 

developed SRP Section 17.5 using the American Society of Mechanical Engineers (ASME) 

Standard NQA-1-1994, “Quality Assurance Requirements for Nuclear Facility Applications,” 

supplemented by additional regulatory and industry guidance for nuclear operating facilities. 

SRP Section 17.5 also addresses additional quality assurance requirements in 10 CFR Part 50, 

Appendix A, GDC 1 and in 10 CFR 50.34(f)(3)(ii) and (iii). GDC 1, “Quality standards and 

records,” requires that a QA Program be established and implemented. 10 CFR 50.34(f)(3)(ii) 

and (iii) specify design and construction QA requirements which must be addressed in a QAPD. 

The introduction of the QAPD, Section 1.1, “Scope/Applicability,” states that, “Portions of the 

QAPD also apply to design quality assurance activities conducted in support of Design 

Certification Aircraft Impact Amendment (AIA). Assessments for AIA are beyond design basis 

and therefore the quality assurance requirements, of 10 CFR Part 50, Appendix B, do not apply. 

Controls in accordance with NEI 07-13 are in place to establish the validity of analyses and 

supporting calculations, and documentation files are maintained with a complete set of 

analyses. Changes to the facility design as a result of the assessments are subject to the full 

design controls described in this QAPD.” 

NRC staff conducted an inspection April 19 through 22, 2010, to provide the staff with 

reasonable assurance that the QA Program has been adequately implemented. The NRC 

staff‘s observations, conclusions, and Notice of Violation (NOV) are documented in Inspection 

Report Nos. 05200012/2010201 and 05200013/2010201 (ML101470298). One violation was 

identified for ineffective corrective action related to the inadequate documentation of personnel 

training and qualification. With the exception of the violation, the NRC inspection team 

concluded that the STPNOC QA policies and procedures complied with the applicable 

requirements of 10 CFR Part 21 and Appendix B to 10 CFR Part 50. The NRC inspection team 

further concluded that STPNOC personnel adequately implemented these policies and 

procedures. Therefore, the staff found that, STPNOC adequately implemented the QA Program 

described in its application as required by 10 CFR 52.54(a)(8). 

17.1.19.4.1 Organization 

The STP Units 3 and 4 QAPD follows the guidance in SRP Section 17.5, paragraph II.A, related 

to organization. The QAPD describes and defines the responsibility and authority for planning, 

establishing, and implementing an effective overall QA Program. The QAPD also describes an 

organizational structure; functional responsibilities; levels of authority; and interfaces for 
17-3

establishing, executing, and verifying implementation of the QAPD. The QAPD establishes 

independence between the organization responsible for verifying a function and the organization 

that performs the function. In addition, the QAPD allows the STP management to size the QA 

organization commensurate with its assigned duties and responsibilities. 

In addition, in the STP Units 3 and 4 QAPD, the applicant commits to comply with the quality 

standards for QA organizations described in NQA-1-1994, “Basic Requirement 1,” and in 

Supplement 1S-1. 

17.1.19.4.2 Quality Assurance Program 

The applicant’s QAPD follows the guidance for the QA Program in SRP Section 17.5, 

paragraph II.B. The QAPD establishes measures that implement a QA Program to ensure that 

the design of a nuclear power plant is in accordance with governing regulations and license 

requirements. The QA Program comprises planned and systematic actions that are necessary 

to provide confidence that the SSCs will perform their intended safety functions, including 

certain non-safety-related SSCs and activities that are significant contributors to plant safety. 

The QA Program requires the maintenance of a list or system that identifies SSCs and activities 

to which the QAPD applies. 

The QAPD provides measures that assess at least once each year or at least once during the 

life of the activity, whichever is shorter, the adequacy of the QAPD and ensure its effective 

implementation. In addition, consistent with SRP Section 17.5 paragraph II.B.8, the QAPD 

applies a grace period of 90 days to activities that must be performed on a periodic basis. The 

next due date for the performance of an activity that invokes the 90-day grace period remains 

unchanged. The next due date for an activity performed before the scheduled due date is 

moved forward, so the interval prescribed for the performance of the activity is not exceeded. 

The QAPD also follows the guidance in SRP Section 17.5, paragraphs II.S and II.T, for training. 

The QAPD describes measures that establish and maintain formal indoctrination and training 

programs for personnel performing, verifying, or maintaining activities within the scope of the 

QAPD to ensure that they achieve and maintain a suitable level of proficiency. The QAPD also 

provides the minimum training requirements for managers responsible for implementing the 

QAPD, in addition to the minimum training requirements for those individuals responsible for 

planning, implementing, and maintaining the QAPD. 

In the QAPD, the applicant commits to comply with the qualification and training standards 

described in NQA-1-1994, Basic Requirement 2; and Supplements 2S-1, 2S-2, 2S-3, and 2S-4, 

with the following clarifications and alternatives: 

• NQA-1-1994, Supplement 2S-1 includes use of the guidance in Appendix 2A-1 the same as 

if it were part of the Supplement. NRC staff evaluated this clarification and determined that 

it is consistent with a previous approval of NEI 06-14, “Quality Assurance Program 

Description,“ Revision 7 (ML092650695) and is therefore acceptable. 

• NQA-1-1994 Supplement 2S-3 states that the prospective lead auditors must have 

participated in a minimum of five audits in the previous 3 years. As an alternative to 

this requirement, the QAPD proposes to follow the guidance in SRP Section 17.5 

paragraph II.S.4.c, which states that the prospective lead auditor shall demonstrate an 

ability to properly conduct the audit process (as implemented by the company), to effectively 

lead an audit team, and to effectively organize and report results, including participation in at 
17-4

least one nuclear audit within the year preceding the date of qualification. NRC staff 

evaluated this proposed alternative and determined that it is consistent with the regulation in 

10 CFR Part 50, Appendix B, Criterion II. Therefore, the staff concluded that this alternative 

is acceptable. 

17.1.19.4.3 Design Control 

The applicant’s QAPD follows the guidance of SRP Section 17.5, paragraph II.C for design 

control. The QAPD establishes the necessary measures that control the design and design 

changes that are subject to the QAPD provisions. The QAPD design process includes 

provisions for controlling design inputs, outputs, changes, interfaces, records, and 

organizational interfaces with the applicant and suppliers. These provisions ensure that the 

design inputs (i.e., design bases and the performance, regulatory, quality, and quality 

verification requirements) are correctly translated into design outputs (i.e., analyses, 

specifications, drawings, procedures, and instructions). In addition, the QAPD provides for 

individuals knowledgeable about QA principles to review design documents to ensure that they 

contain the necessary QA requirements. 

In the QAPD, the applicant commits to comply with the quality standards described in 

NQA-1-1994, Basic Requirement 3 and Supplement 3S-1, to establish the program for design 

control and verification. The applicant also commits to comply with the requirements of 

Subpart 2.20 for the subsurface investigation requirements and Subpart 2.7 for the standards 

for computer software QA controls. NRC staff found these commitments acceptable. 

17.1.19.4.4 Procurement Document Control 

The applicant’s QAPD follows the guidance in SRP Section 17.5 paragraph II.D, for 

procurement document control. The QAPD establishes the necessary administrative controls 

and processes to ensure that procurement documents include or reference applicable 

regulatory, technical, and QA Program requirements. As noted in SRP Section 17.5 

paragraph II.D.1, the applicable technical, regulatory, administrative, quality, and reporting 

requirements (such as specifications, codes, standards, tests, inspections, special processes, 

and the regulation in 10 CFR Part 21, ”Reporting of Defects and Noncompliance,”) are invoked 

for the procurement of items and services. 

In the QAPD, the applicant commits to comply with the quality standards described in 

NQA-1-1994, Basic Requirement 4 and Supplement 4S-1, with the following alternatives and 

commitment. 

• NQA-1-1994 Supplement 4S-1, Section 2.3 states that procurement documents must 

require suppliers to have a documented QA Program that implements NQA-1-1994, Part I. 

– As an alternative to this requirement, the QAPD proposes that suppliers have a 

documented QA Program that meets the requirements of Appendix B to 10 CFR Part 50, 

as applicable to the circumstances of the procurement. NRC staff evaluated this 

proposed alternative and determined that it is consistent with Appendix B, Criterion IV, 

“Procurement Document Control.” Therefore, the staff concluded that this alternative is 

acceptable. 

– As an alternative to this requirement, the QAPD proposes that procurement documents 

allow suppliers to work under the applicant’s QAPD, including implementing procedures, 

if suppliers do not have their own QA Program. NRC staff evaluated this proposed 
17-5

alternative and determined that the applicant’s QAPD follows the guidance in SRP 

Section 17.5 paragraph II.G, “Control of Purchased Material, Equipment, and Services.” 

Specifically, the QAPD provides measures to evaluate prospective suppliers so that only 

qualified suppliers are selected, acceptance actions are performed for procured products 

and services, and suppliers are periodically audited and evaluated to ensure that 

qualified suppliers continue to provide acceptable products and services. Therefore, the 

staff concluded that this alternative is acceptable. 

• NQA-1-1994 Supplement 4S-1, Section 3 states that procurement documents are to be 

reviewed before awarding the contract. As an alternative to this requirement, the QAPD 

proposes to conduct the QA review of procurement documents through the review of the 

applicable procurement specifications, including the technical and quality procurement 

requirements, before awarding the contract. In addition, procurement document changes 

(e.g., scope, technical, or quality requirements) will also receive a QA review. NRC staff 

evaluated this proposed alternative and determined that it provides an adequate QA review 

of procurement documents before awarding the contract and after any change. Therefore, 

the staff concluded that this alternative is acceptable. 

• In the QAPD, the applicant commits to ensure that procurement documents prepared for 

commercial-grade items and procured for use as safety-related items shall contain technical 

and quality requirements, so that the procured item can be appropriately dedicated. NRC 

staff evaluated this proposed commitment and determined that it is consistent with staff 

guidance in Generic Letter 89-02, “Actions to Improve the Detection of Counterfeit and 

Fraudulently Marked Products,” dated March 21, 1989 (Reference 3); and Generic 

Letter 91-05, “Licensee Commercial-Grade Procurement and Dedication Programs,” dated 

April 9, 1991 (Reference 4); as delineated in SRP Section 17.5, paragraphs II.U.1.d and 

II.U.1.e. Therefore, the staff concluded that this commitment is acceptable. 

17.1.19.4.5 Instructions, Procedures, and Drawings 

The applicant’s QAPD follows the guidance in SRP Section 17.5 paragraph II.E for instructions, 

procedures, and drawings. The QAPD establishes the necessary measures and governing 

procedures to ensure that activities affecting quality are prescribed by and performed in 

accordance with documented instructions, procedures, and drawings. 

In the QAPD, the applicant commits to comply with the quality standards for instructions, 

procedures, and drawings described in NQA-1-1994, Basic Requirement 5 to establish 

procedural controls. 

17.1.19.4.6 Document Control 

The applicant’s QAPD follows the guidance of SRP Section 17.5, paragraph II.F for document 

control. The QAPD establishes the necessary measures and governing procedures to control 

the preparation, review, approval, issuance, and revision of documents that specify quality 

requirements or prescribe measures for controlling activities affecting quality, including 

organizational interfaces. The QAPD provides measures to ensure that the same organization 

that performed the original review and approval should also review and approve revisions or 

changes to documents, unless other organizations are specifically designated. A listing of all 

controlled documents identifying the currently approved revision or date is maintained so 

personnel can readily determine the appropriate document for use. 
17-6

In the QAPD, the applicant commits to comply with the quality standards for document control 

described in NQA-1-1994, Basic Requirement 6 and Supplement 6S-1, to establish provisions 

for document control. 

17.1.19.4.7 Control of Purchased Material, Equipment, and Services 

The applicant’s QAPD follows the guidance in SRP Section 17.5, paragraph II.G for the control 

of purchased material, equipment, and services. The QAPD establishes the necessary 

measures and governing procedures to control the procurement of items and services that 

ensure conformance to specified requirements. The program provides measures to evaluate 

prospective suppliers so that only qualified suppliers are selected. In addition, the program 

requires the suppliers to be periodically audited and evaluated to ensure that qualified suppliers 

continue to provide acceptable products and services. 

The program provides acceptance actions such as source verification, receipt inspection, 

pre-and post-installation tests, and the review of documentation such as certificates of 

conformance to ensure that procurement, inspection, and testing requirements have been 

satisfied before relying on the item to perform its intended safety function. Purchased items and 

services are subject to quality and technical requirements at least equivalent to those specified 

for original equipment—or by properly reviewed and approved revisions—to ensure that the 

items are suitable for the intended service and are of acceptable quality, consistent with their 

effects on safety. 

In the QAPD, the applicant commits to comply with the quality standards for the control of 

purchased material, equipment, and services described in NQA-1-1994, Basic Requirement 7 

and Supplement 7S-1, to establish procurement verification control with the following exceptions 

and alternatives. 

• NQA-1-1994, Basic Requirement 7 and Supplement 7S-1, states that procurement sources 

and the performance of suppliers are to be evaluated. As an exception to these 

requirements, the QAPD proposes that other 10 CFR Part 50 licensees (other than the 

STP), authorized nuclear inspection agencies, the National Institute of Standards and 

Technology (NIST), and other State and Federal agencies that may provide items or 

services to STP are not required to be evaluated or audited. 

NRC staff acknowledged that 10 CFR Part 50 licensees, authorized nuclear inspection 

agencies, the National Voluntary Laboratory Accreditation Program (NVLAP) administered 

by NIST, and other State and Federal agencies perform work under quality programs 

acceptable to the NRC and no additional audits or evaluations are required. However, 

STPNOC remains responsible for ensuring that procured items or services conform to the 

Appendix B program, the applicable ASME Boiler and Pressure Vessel Code requirements, 

and other regulatory requirements and commitments. The applicant also remains 

responsible for ensuring that items or services are suitable for the intended application and 

for documenting the evaluation that supports this conclusion. The staff determined that this 

proposed exception provides an appropriate level of quality and safety and is therefore 

acceptable, as documented in a previous SER (ML003693241). 

• SRP Section 17.5, paragraph II.L.8 establishes provisions for the procurement of 

commercial-grade calibration services for safety-related applications. As an exception to 

these provisions, the QAPD proposes not to require procurement source evaluation and 

selection measures, provided that all of the following conditions are met: 
17-7

– Purchase documents will impose additional technical and administrative requirements to 

satisfy QAPD and technical requirements. 

– Purchase documents will require the reporting of as-found calibration data, when 

calibrated items are found to be out of tolerance. 

– A documented review of the supplier’s accreditation will be performed and will include a 

verification of the following: 

(1) The calibration laboratory holds a domestic accreditation by any one of the following 

accrediting bodies, which are recognized by the International Laboratory 

Accreditation Cooperation (ILAC) Mutual Recognition Arrangement (MRA): 

a. NVLAP, administered by NIST 

b. American Association for Laboratory Accreditation (A2LA) 

c. ACLASS Accreditation Services (ACLASS) 

d. International Accreditation Service (IAS) 

e. Laboratory Accreditation Bureau (L-A-B) 

(2) The accreditation encompasses American Nuclear Society/International 

Standardization Organization/International Electrotechnical Commission 17025, 

“General Requirements for the Competence of Testing and Calibration Laboratories." 

(3) The published scope of accreditation for the calibration laboratory covers the 

necessary measurement parameters, ranges, and uncertainties. 

NRC staff evaluated the NVLAP and A2LA accreditation programs and found them both 

acceptable (ML052710224). The staff subsequently determined that the accreditation of the 

ACLASS, L-A-B, and IAS Programs is also recognized by the ILAC MRA and is therefore 

acceptable (ML073440472, ML081140564, and ML081330253). 

• NQA-1-1994 Supplement 7S-1, Section 8.1 states that documentary evidence that items 

conform to procurement documents shall be available at the nuclear facility site before 

installation or use. As an alternative to the requirement for procurement documentary 

evidence to be available at the nuclear facility site during construction, the QAPD proposes 

that documentary evidence may be stored in physical form or in electronic media, under the 

control of STP or its supplier, at a location other than the nuclear facility site, as long as the 

documents can be accessed at the nuclear facility site during construction. After the 

construction is complete, sufficient documentary evidence will be available to the licensee to 

support operations. 

The staff determined that implementation of this alternative would allow access to and review of 

the necessary procurement documentary evidence at the nuclear facility site, both before 

installation and after use. Therefore, the staff concluded that this alternative is acceptable. 

• As an alternative to the requirements for the control of commercial-grade items and services 

in NQA-1-1994 Supplement 7S-1 Section 10, the applicant commits in the QAPD to follow 

NRC guidance discussed in Generic Letters 89-02 and 91-05. In SRP Section 17.5, 

paragraphs II.U.1.d and II.U.1.e provide guidance to establish and describe special quality 

verification requirements in applicable documents to ensure that the commercially procured 

items will perform satisfactorily in service. In addition, the documents should provide for 

determining critical characteristics, technical evaluations, receipt requirements, and quality 

evaluations of the items to ensure that the items are suitable for their intended use. 
17-8

The staff determined that this alternative will improve the likelihood of detecting counterfeit and 

fraudulently marked products and will improve the commercial-grade dedication programs. 

Therefore, the staff concluded that this alternative is acceptable. 

17.1.19.4.8 Identification and Control of Materials, Parts, and Components 

This element is not applicable to the DC amendment application and has not been reviewed or 

approved by the NRC staff. 

17.1.19.4.9 Control of Special Processes 

This element is not applicable to the DC amendment application and has not been reviewed or 

approved by the NRC staff. 

17.1.19.4.10 Inspection 

The applicant’s QAPD follows the guidance in SRP Section 17.5, paragraph II.J for inspections. 

The QAPD establishes the necessary measures to implement inspections to ensure that items, 

services, and activities affecting safety meet established requirements and conform to 

applicable documented specifications, instructions, procedures, and design documents. The 

inspection program establishes requirements for planning inspections, determining applicable 

acceptance criteria, setting the frequency of inspections, and identifying special tools needed to 

perform the inspection. Properly qualified personnel independent of those who perform or 

directly supervise the work are required to perform the inspections. 

In the QAPD, the applicant commits to comply with the quality standards for inspections 

described in NQA-1-1994, Basic Requirement 10, Supplement 10S-1, and Subparts 2.4, 2.5, 

and 2.8 to establish inspection requirements with the following commitment and alternative: 

• NQA-1-1994, Subpart 2.4 requires the use of the IEEE Std 336-1985, “IEEE Standard 

Installation, Inspection, and Testing Requirements for Power, Instrumentation, and Control 

Equipment at Nuclear Facilities.” IEEE Std 336-1985 refers to IEEE Std 498-1985, “IEEE 

Standard Requirements for the Calibration and Control of Measuring and Test Equipment 

Used in Nuclear Facilities.” Each of these standards uses the definition of safety systems 

equipment from IEEE Std 603-1980, “IEEE Standard Criteria for Safety Systems for Nuclear 

Power Generating Stations.” IEEE Std 603-1980 defines “safety system” as 

Those systems (the reactor trip system, an engineered safety feature, or both, 

including all their auxiliary supporting features and other auxiliary feature) which 

provide a safety function. A safety system is comprised of more than one safety 

group of which any one safety group can provide the safety function. 

In the QAPD, the applicant must commit to the definition of safety systems equipment from 

IEEE Std 603-1980 to appropriately implement NQA-1-1994, Subpart 2.4. In the QAPD, the 

applicant commits to the definition of safety systems equipment from IEEE Std 603-1980 but 

does not commit to the balance of IEEE Std 603-1980. This definition applies only to 

equipment in the context of NQA-1-1994, Subpart 2.4. NRC staff determined that the use of 

the definition of safety systems equipment is acceptable because it is consistent with the 

requirements of NQA-1-1994, Subpart 2.4. 
17-9

17.1.19.4.11 Test Control 

The applicant’s QAPD follows the guidance in SRP Section 17.5, paragraph II.K for test control. 

The QAPD establishes the necessary measures and governing provisions to demonstrate that 

items subject to the provisions of the QAPD will perform satisfactorily in service; that the plant 

can be operated safely as designed; and that the operation of the plant, as a whole, is 

satisfactory. 

In the QAPD, the applicant commits to comply with the quality standards for test control 

described in NQA-1-1994, Basic Requirement 11, and Supplement 11S-1 to establish 

provisions for testing. 

Furthermore, in the QAPD, the applicant commits to comply with the quality standards for 

software test control described in NQA-1-1994, Supplement 11S-2 and Subpart 2.7 to establish 

provisions ensuring that computer software used in applications affecting safety will be 

prepared, documented, verified, tested, and used so that the expected outputs are obtained and 

the configuration control is maintained. 

17.1.19.4.12 Control of Measuring and Test Equipment 

The applicant’s QAPD follows the guidance in SRP Section 17.5, paragraph II.L for the control 

of measuring and test equipment (M&TE). The QAPD establishes the necessary measures to 

control the calibration, maintenance, and use of M&TE that provide information important to a 

safe plant operation. 

In the QAPD, the applicant commits to comply with the quality standards for M&TE described in 

NQA-1-1994, Basic Requirement 12, and Supplement 12S-1 to establish provisions for control 

of M&TE with the following clarification and exception: 

• The QAPD clarifies that the out-of-calibration conditions described in paragraph 3.2 of 

Supplement 12S-1 of NQA-1-1994 refer to cases where the M&TE are found to be out of the 

required accuracy limits (i.e., out of tolerance) during calibration. NRC staff determined that 

the clarification for the out-of-calibration conditions is consistent with Supplement 12S-1. 

Therefore, the staff concluded that this clarification is acceptable. 

• As an alternative to NQA-1-1994 Subpart 2.4 Section 7.2.1, “Calibration Labeling 

Requirements,” the QAPD proposes that when it is impossible or impractical to mark 

equipment with the required calibration information because of equipment size or 

configuration, the required calibration information will be documented and traceable to the 

equipment. NRC staff determined that this alternative is consistent with the guidance in 

SRP 17.5, paragraph II.L.3. Therefore, the staff concluded that this alternative is 

acceptable. 

17.1.19.4.13 Handling, Storage, and Shipping 

This element is not applicable to the DC amendment application and has not been reviewed or 

approved by the NRC staff. 

17.1.19.4.14 Inspection, Test, and Operating Status 

This element is not applicable to the DC amendment application and has not been reviewed or 

approved by the staff. 
17-10

17.1.19.4.15 Nonconforming Materials, Parts, or Components 

The applicant’s QAPD follows the guidance in SRP Section 17.5, paragraph II.O for 

nonconforming materials, parts, or components. The QAPD establishes the necessary 

measures to control items, including services, that do not conform to specified requirements to 

prevent inadvertent installation or use. Nonconformances are evaluated for their impact on the 

operability of quality SSCs to ensure that the final condition does not adversely affect the safety, 

operation, or maintenance of the item or service. The results of evaluations of conditions that 

adversely affect quality are analyzed to identify quality trends documented and reported to 

upper management, in accordance with the applicable procedures. 

In addition, the QAPD establishes the necessary measures to implement the requirements of 

10 CFR Part 52, 10 CFR 50.55, and 10 CFR Part 21, as applicable. 

In the QAPD, the applicant commits to comply with the standards of quality for nonconforming 

materials, parts, or components described in NQA-1-1994, Basic Requirement 15, and 

Supplement 15S-1 to establish measures for nonconforming materials. 

17.1.19.4.16 Corrective Action 

The applicant’s QAPD follows the guidance in SRP Section 17.5 paragraph II.P, for Corrective 

Action Programs. The QAPD establishes the necessary measures to promptly identify, control, 

document, classify, and correct conditions that adversely affect quality. The QAPD requires 

personnel to identify known conditions that adversely affect quality. Reports of conditions that 

adversely affect quality are analyzed to identify trends. Significant conditions that adversely 

affect quality are documented and reported to responsible management. In the case of 

suppliers working on safety-related activities or similar situations, the applicant may delegate 

specific responsibilities for the Corrective Action Program, but the applicant maintains 

responsibility for the program's effectiveness. 

In addition, the QAPD establishes the necessary measures to implement the requirements of 

10 CFR Part 52, 10 CFR 50.55, and 10 CFR Part 21, as applicable. 

In the QAPD, the applicant commits to comply with the standards of quality for corrective 

actions described in NQA-1-1994 Basic Requirement 16 to establish a Corrective Action 

Program. 

17.1.19.4.17 Quality Assurance Records 

The applicant’s QAPD follows the guidance in SRP Section 17.5, paragraph II.Q for QA 

records. The QAPD establishes the necessary measures to ensure that sufficient records of 

items and activities affecting quality are generated, identified, retained, maintained, and able to 

be retrieved. 

In establishing measures to ensure that sufficient records of completed items and activities 

affecting quality are appropriately stored, the QAPD commits Regulatory Position C.2 and 

Table 1 of RG 1.28, Revision 3, for records and retention times. 

Concerning the use of storage and retrieval systems for electronic records, the QAPD complies 

with the NRC guidance in Generic Letter 88-18, "Proposed Final NRC Generic Letter 88-18, 

Supplement 1,” “Guidance on Managing Quality Assurance Records in Electronic Media,” 
17-11

dated September 13, 1999; Regulatory Issue Summary 2000-18, "Guidance on Managing 

Quality Assurance Records in Electronic Media,” dated October 23, 2000; and associated 

Nuclear Information and Records Management Association (NIRMA) guidelines TG 11-1998, 

TG 15-1998, TG 16-1998, and TG 21-1998. 

In the QAPD, the applicant commits to comply with the standards for quality of QA records 

described in NQA-1-1994, Basic Requirement 17, and Supplement 17S-1 to establish 

provisions for records with the following alternative: 

• In NQA-1-1994 Supplement 17S-1, Section 4.2(b) states that records must be firmly 

attached in binders or placed in folders or envelopes for storage in steel file cabinets or on 

shelving in containers. As an alternative to this requirement, the QAPD proposes that 

hard-copy records be stored in steel cabinets or on shelving in containers, except that 

methods other than binders, folders, or envelopes may be used to organize records for 

storage. 

NRC staff determined that this alternative is acceptable, as documented in a previous SER 

(ML052360625). 

17.1.19.4.18 Quality Assurance Audits 

The applicant’s QAPD follows the guidance of SRP Section 17.5, paragraph II.R for QA audits. 

The QAPD establishes the necessary measures to implement audits to verify that activities 

covered by the QAPD are performed in conformance with documented requirements. The Audit 

Program is reviewed for effectiveness as part of the overall audit process. 

In the QAPD, the applicant provides for conducting periodic internal and external audits. 

Internal audits are conducted to determine the adequacy of program and procedures, and to 

determine if they are meaningful and comply with the overall QAP. Internal audits are 

performed with a frequency to ensure that an audit of all applicable QA program elements is 

completed within a period of once per calendar year or at least once during the life of the 

activity, whichever is shorter. External audits determine the adequacy of a supplier’s or 

contractor’s QA Program. 

The applicant ensures that audits are documented and audit results are reviewed. In 

accordance with the QAPD, the applicant will respond to all audit findings and initiate 

appropriate corrective actions. In addition, where corrective actions are indicated, the applicant 

documents the follow-up of applicable areas through inspections, reviews, re-audits, or other 

appropriate means to verify the implementation of assigned corrective actions. 

In the QAPD, the applicant commits to comply with the quality standards for QA audits 

described in NQA-1-1994, Basic Requirement 18, and Supplement 18S-1 to establish an 

independent Audit Program. 

17.1.19.4.19 Non-Safety-Related SSC Quality Assurance Controls 

17.1.19.4.19.1 Non-Safety-Related SSCs – Significant Contributors to Plant Safety 

The applicant’s QAPD follows the guidance of SRP Section 17.5, paragraph II.V.1 on controls 

related to nonsafety-related SSCs. The QAPD establishes program controls applied to 

nonsafety-related SSCs that are significant contributors to plant safety but are not covered by 
17-12

Appendix B. The QAPD applies specific controls to these items in a selected manner, targeting 

the characteristics or critical attributes that render the SSC a significant contributor to plant 

safety consistent with applicable sections of the QAPD. 

17.1.19.4.19.2 Non-Safety-Related SSCs Credited for Regulatory Events 

The applicant’s QAPD follows the guidance of SRP Section 17.5, paragraph II.V.2 to establish 

the quality requirements for non-safety-related SSCs credited for regulatory events. In the 

QAPD, the applicant commits to comply with the following regulatory guidance: 

• The applicant shall implement quality provisions for the fire protection system in accordance 

with Regulatory Position 1.7, “Quality Assurance,” in RG 1.189, “Fire Protection for 

Operating Nuclear Power Plants,” issued April 2001. 

• The applicant shall implement quality provisions for anticipated transient without scram 

(ATWS) equipment in accordance with Generic Letter 85-06, “Quality Assurance Guidance 

for ATWS Equipment That Is Not Safety Related,” issued January 1985. 

• The applicant shall implement quality provisions for station blackout (SBO) equipment in 

accordance with Regulatory Position 3.5, “Quality Assurance and Specific Guidance for 

SBO Equipment That Is Not Safety Related,” and Appendix A, “Quality Assurance Guidance 

for Non-Safety Systems and Equipment,” in RG 1.155, “Station Blackout,” issued 

August 1988. 

17.1.19.4.20 Regulatory Commitments 

The applicant’s QAPD follows the guidance of SRP Section 17.5, paragraph II.U for describing 

regulatory commitments. The QAPD establishes QA Program commitments. In the QAPD, the 

applicant commits to comply with the following NRC regulatory guides and other QA standards 

to supplement and support the QAPD, with the noted clarifications and alternatives. 

• RG 1.26, Revision 4, “Quality Group Classification and Standards for Water-, Steam, and 

Radioactive-Waste-Containing Components of Nuclear Power Plants,” issued March 2007. 

The QAPD states that STPNOC conforms to the applicable regulatory positions through 

DCD Section 3.2. Chapter 3 of this SER includes additional details. NRC staff reviewed this 

clarification and found it consistent with the guidance in SRP 17.5 and therefore acceptable. 

• RG 1.28 Revision 3, “Quality Assurance Program Requirements (Design and Construction),” 

issued August 1985. 

The QAPD identifies an alternative to Regulatory Position C.1 in Section 2.7 as accepted in 

a previous SER (ML070510300). The applicant states that Regulatory Positions C.3.1 and 

C.3.2 are addressed in Sections 18.2 and 7.1 of the QAPD, respectively. The staff reviewed 

these clarifications and alternatives and found them consistent with the guidance in 

SRP 17.5 and therefore acceptable. 

The QAPD states that Regulatory Position C.2 is addressed in Section 17.1 

• RG 1.29 Revision 4, “Seismic Design Classification,” issued March 2007. 

The QAPD states that STPNOC conforms to the applicable Regulatory Positions C.1 

through C.3 through DCD Section 3.2. Chapter 3 of this SER includes additional details. 
17-13

The QAPD states that Regulatory Position C.4 is addressed by the QAPD requirements for 

safety-related activities. The staff reviewed these clarifications and found them consistent 

with the guidance in SRP 17.5 and therefore acceptable. Regulatory Position C.5 does not 

apply to AIA amendment activities. 

• ASME NQA-1-1994, “Quality Assurance Requirements for Nuclear Facility Applications,” 

Parts I and II, as described in foregoing sections. 

• NIRMA technical guides, as described in Subsection 17.1.19.17 of the QAPD. 

17.1.19.4.21 Additional Quality Assurance and Administrative Controls for the Plant 

Operational Phase 

This element is not applicable to the DC amendment application and has not been reviewed or 

approved by the staff. 

17.1.19.5 Conclusion 

NRC staff reviewed the QAP of the STPNOC's application to amend the Design Certification 

Rule for the ABWR. The staff reviewed the proposed Section 17.1.19 of the application. 

Section 17.1.19 references the description of the QA Program applied to the DCD amendment 

activities associated with the STPNOC aircraft impact assessment using the STP Units 3 and 4 

QAPD. The staff based the review of the STP Units 3 and 4 QAPD, Revision 4, on 

10 CFR 52.47(a)(19); 10 CFR 52.54(a)(8); 10 CFR Part 50, Appendix B, “Quality Assurance 

Criteria for Nuclear Power Plants and Fuel Reprocessing Plants”; and SRP Section 17.5. The 

staff found that the QAPD used for the design phase of the proposed amendment to the ABWR 

Design Certification Rule is acceptable. 

On the basis of the NRC staff’s review of the STP Units 3 and 4 QAPD, the NRC staff 

concluded that: 

• The STP Units 3 and 4 QAPD adequately describes the authority and responsibility of 

management and supervisory personnel, performance/verification personnel, and audit 

personnel. 

• The STP Units 3 and 4 QAPD adequately provides for organizations and persons 

responsible for performing the verification and audit functions have the authority and 

independence to conduct their activities without undue influence from those directly 

responsible for costs and schedules. 

• The STP Units 3 and 4 QAPD adequately applies to activities and items that are important 

to safety. 

• The STP Units 3 and 4 QAPD adequately describes the program for the QA treatment of 

non-safety-related SSCs. 

• The STP Units 3 and 4 QAPD adequately describes a philosophy and controls that, 

when properly implemented, comply with the requirements of 10 CFR 50.34(f)(3)(ii) and 

(iii) pursuant to 10 CFR 52.47(a)(8); Appendix B to 10 CFR Part 50 pursuant to 

10 CFR 52.47(a)(19); and GDC 1. 
17-14

Therefore, the NRC staff concluded that the STP Units 3 and 4 QAPD adequately describes the 

STPNOC’s QA Program. Accordingly, the staff concluded that the STP Units 3 and 4 QAPD 

complies with the applicable NRC regulations and industry standards and can be used by 

STPNOC for DC amendment activities associated with the ABWR. The staff also concluded 

that the applicant adequately implemented the QA Program described in its application as 

required by 10 CFR 52.54(a)(8).
19S-1

19S DESIGN FEATURES FOR PROTECTION AGAINST A 

LARGE, COMMERCIAL AIRCRAFT IMPACT 

19S.1 Introduction 

As part of the application, STPNOC adds Appendix 19S to provide design features for 

protection against a large commercial aircraft impact. This section of the SER describes the 

staff’s evaluation of the description of design features and functional capabilities credited by the 

applicant to show that the facility can withstand the effects of a large, commercial aircraft 

impact. 

The impact of a large commercial aircraft is a beyond-design-basis event. Under 

10 CFR 50.150, applicants for new nuclear power reactors1

 are required to perform an 

assessment of the effects on the designed facility of the impact of a large commercial aircraft. 

Applicants are required to submit a description of the design features and functional capabilities 

identified as a result of the assessment (key design features) in their DCD, together with a 

description of how the identified design features and functional capabilities show that the 

acceptance criteria in 10 CFR 50.150(a)(1) are met. Applicants subject to 10 CFR 50.150 must 

make the complete aircraft impact assessment available for NRC inspection at the applicants’ 

offices or their contractors’ offices, upon an NRC request in accordance with 10 CFR 50.70, 

10 CFR 50.71, and Section 161.c of the Atomic Energy Act of 1954, as amended. 

19S.2 Summary of Application 

In DCD Tier 2, Revision 3, Appendix 19S, the applicant states that an aircraft impact 

assessment in accordance with the requirements in 10 CFR 50.150(a)(1) was performed using 

the methodology described in NEI 07-13, “Methodology for Performing Aircraft Impact 

Assessments for New Plant Designs,” Revision 7, as endorsed by the NRC in the draft 

regulatory guide, DG-1176. Based on the results of the assessment, the applicant has identified 

a set of key design features to show that the acceptance criteria in 10 CFR 50.150(a)(1) are 

met. In the July 12, 2010, AIA amendment submittal letter, the applicant states “all assumptions 

from the assessment documents, without which in our opinion the success criteria would not be 

met, have been included as key design features.” These key design features are reported in 

DCD Tier 2, Revision 3, Appendix 19S along with references to other sections of the DCD that 

provide additional details. DCD Tier 2, Revision 3, Appendix 19S also contains descriptions of 

how the key design features show that the acceptance criteria in 10 CFR 50.150(a)(1) are met. 

In the initial DCD Amendment Table 1.9-1, “Summary of ABWR Standard Plant COL License 

Information,” the applicant identified the following COL license information items: 

ITEM No. SUBJECT

19.9k Procedures for Use of Alternate Feedwater Injection

19.9l Procedures to Depressurize the Reactor Pressure Vessel (RPV) from the AFI 

Pump House

19.9m Verification of Environmental Conditions in the AFI Pump House

19.9n Description of Electrical Power Supply for AFI Equipment.

 

1

 “Applicants for new nuclear power reactors” is defined in the Statement of Considerations for the Aircraft 

Impact Rule (74 FR 28112, June 12, 2009).
19S-2

In the DCD amendment dated May 12, 2010 (ML101340548), the applicant deleted the COL 

license information items listed above, since information was provided in the amended ABWR 

DCD as ITAAC. 

19S.2.1 Description of Key Design Features 

The credited key design features, their function(s), and references to sections containing the 

detailed descriptions are summarized below: 

• The primary containment structure, as described in DCD Tier 2, Revision 3, Sections 3.8 

and 3H.1, and Figure 1.2-1 protects the safety systems inside from impact. 

• The location and design of the control building structure, as described in DCD Tier 2, 

Revision 3, Sections 1.2, 3.8.4, and 3H.2 and Figure 1.2-1 protects the north wall of the RB 

from impact. 

• The location and design of the turbine building structure, as described in DCD Tier 1, 

Revision 3, Section 2.15.11 and Tier 2, Section 1.2 with Figures 1.2-1 and 1.2-24 through 

1.2-31, protects the north wall of the control building and reactor building from impact. 

• The location and design of the reactor building structure, as described in DCD Tier 2, 

Revision 3, Sections 1.2, 3.8.4, and 3H.1 and Figure 1.2-1 protects the south wall of the CB 

and primary containment from impact. 

• The location and design of the spent fuel pool (SFP) and its supporting structure as 

described in DCD Tier 2, Revision 3, Sections 1.2 and 9.1 and Figure 1.2-12, protect the 

spent fuel pool from impact. 

• The physical separation of the Class 1E emergency diesel generators and an independent 

power supply as described in DCD Tier 2, Revision 3, Section 9.5.14, prevent the loss of all 

electrical power to core cooling systems. 

• The location and design of 3-hour fire barriers, including fire doors and watertight doors 

inside the reactor building and control building, as described in DCD Tier 2, Revision 3, 

Sections 9.5.1 and 9A.4 protect credited core cooling equipment from fire damage. 

• The physical separation and design of the ECCS, as described in DCD Tier 2, Revision 3, 

Section 6.3 ensure core cooling. 

• The design of the AFI system as described in DCD Tier 2, Revision 3, Section 9.5.14 

ensures core cooling. 

• The design of the containment overpressure protection system (COPS) as described in DCD 

Tier 2, Revision 3, Section 6.2.5 ensures core cooling. 

19S.2.2 Description of How Regulatory Acceptance Criteria are Met 

The acceptance criteria in 10 CFR 50.150(a)(1) are (1) the reactor core will remain cooled or the 

containment will remain intact, and (2) spent fuel pool cooling or spent fuel pool integrity is 

maintained. The applicant states that it has met 10 CFR 50.150(a)(1) by maintaining both core 

cooling and spent fuel pool integrity. 
19S-3

As indicated in the amended ABWR DCD Tier 2, Revision 3, Appendix 19S, the applicant 

proposes to maintain core cooling using the safety-related and non-safety-related systems 

described in DCD Tier 2, Revision 3, Appendix 19S, which are specifically designed to ensure 

that the reactor can be shutdown and decay heat can be removed adequately from the reactor 

core. Some of this equipment is located (1) inside of the primary containment, (2) inside the 

reactor building, and (3) well away from the power block. Locations inside the primary 

containment are protected from structural, shock and fire damage by the design of the primary 

containment structure as well as the reactor building structure that limits the penetration of a 

large, commercial aircraft so that the primary containment is not perforated. Equipment inside 

the reactor building is protected by structural design features of the reactor building itself and by 

structures adjacent to the reactor building, including the turbine building and the control building. 

In addition, fire barriers are designed and located in the reactor building and control building to 

limit the spread of fire inside the buildings. 

The applicant proposes to satisfy the spent fuel pool integrity acceptance criterion in 

10 CFR 50.150(a)(1) due to the location and design of the spent fuel pool and its support 

structure. These key design features protect the structure from impact by a large commercial 

aircraft. 

19S.2.3 Inspection, Test, Analysis, and Acceptance Criteria (ITAAC) 

In accordance with 10 CFR 52.47(b)(1), the applicant has proposed the following ITAAC, as 

described in the amended ABWR DCD Tier 1, Revision 3, Section 2.11.24 and Table 2.11.24 

(ML101340548) for key design features credited to meet 10 CFR 50.150: 

ITAAC# 2.11.24-1– The basic configuration of the AFI system is as described in 

Section 2.11.24. 

ITAAC# 2.11.24-2– The AFI pump is capable of injecting ≥800gpm into the RPV at the 

lowest SRV safety lift pressure 

ITAAC# 2.11.24-3– The AFI system water supply has a minimum capacity of 

300,000 gallons and is refillable. 

ITAAC# 2.11.24-4– The AFI pump house is located a minimum of 300 feet from the 

nearest outside wall of each of the reactor building, control building and turbine building. 

ITAAC# 2.11.24-5– The AFI water supply is located a minimum of 300 feet from the 

nearest outside wall of each of the reactor building, control building and turbine building. 

ITAAC# 2.11.24-6– The AFI power supply is located a minimum of 300 feet from the 

nearest outside wall of each of the reactor building, control building and turbine building. 

ITAAC# 2.11.24-7– Barriers exist, which qualify as the intervening structure as defined in 

NEI 07-13, Revision 7, between the AFI pump house and each of the reactor building, 

control building and turbine building. 

ITAAC# 2.11.24-8– Barriers exist, which qualify as the intervening structure as defined in 

NEI 07-13, Revision 7, between the AFI water supply and each of the reactor building, 

control building and turbine building. 
19S-4

ITAAC# 2.11.24-9– Barriers exist, which qualify as the intervening structure as defined in 

NEI 07-13, Revision 7, between the AFI power supply and each of the reactor building, 

control building and turbine building. 

ITAAC# 2.11.24-10– Instrumentation exists to provide information to the operator in the 

AFI pump house for reactor vessel water level, reactor pressure, suppression pool water 

level, and wetwell pressure. 

ITAAC# 2.11.24-11– MOVs in the AFI system injection line operate as designed on a 

manual initiation signal. 

19S.3 Regulatory Basis 

NRC staff used the following relevant regulations and regulatory guidance to perform this 

review: 

• 10 CFR 50.150(a)(1) requires that applicants perform a design specific assessment of the 

effects on the facility of the impact of a large, commercial aircraft. Using realistic analyses, 

the applicant shall identify and incorporate into the design those design features and 

functional capabilities to show that, with reduced use of operator actions: (i) The reactor 

core remains cooled, or the containment remains intact; and (ii) spent fuel cooling or spent 

fuel pool integrity is maintained. 

• 10 CFR 50.150(b) requires that the final safety analysis report include a description of: 

(1) the design features and functional capabilities which the applicant has identified for 

inclusion in the design to show that the facility can withstand the effects of a large, 

commercial aircraft impact in accordance with 10 CFR 50.150(a)(1); and (2) how those 

design features and functional capabilities meet the assessment requirements of 

10 CFR 50.150(a)(1). 

• 10 CFR 52.47(b)(1) requires that a DC application contain the proposed ITAAC that are 

necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and 

analyses are performed and the acceptance criteria met, a plant that incorporates the 

design certification has been constructed, and will operate in accordance with the design 

certification, the provisions of the Atomic Energy Act, and the NRC’s regulations. 

19S.3.1 Review Guidance 

• DG-1176 “Guidance for the Assessment of Beyond-Design-Basis Aircraft Impacts,” issued 

July 2009, provides guidance for meeting the requirements in 10 CFR 50.150(a), and 

specifically, documents NRC endorsement of the methodologies described in the industry 

guidance document, NEI 07-13, “Methodology for Performing Aircraft Impact Assessments 

for New Plant Designs,” Revision 7, issued May 2009. 

• Statements of Consideration for the “Final Rule on Consideration of Aircraft Impacts for New 

Nuclear Power Reactors” (74 FR 28112, June 12, 2009) which indicates that for the NRC to 

conclude that the rule has been met, it must find that the applicant has performed an aircraft 

impact assessment reasonably formulated to identify design features and functional 

capabilities to show, with reduced use of operator action, that the acceptance criteria in 

10 CFR 50.150(a)(1) are met. 
19S-5

Reasonably Formulated Assessment 

The NRC considers an aircraft impact assessment performed by qualified personnel 

using a method that conforms to the guidance in NEI 07-13, Revision 7 to be a method 

that is reasonably formulated. The NRC considers qualified personnel to be (1) an 

applicant who is the designer of the facility for which the aircraft impact assessment 

applies; or (2) an applicant’s primary contractor for the aircraft impact assessment who 

has designed a nuclear power reactor facility either already licensed or certified by the 

NRC or currently under review by the NRC. 

Cooling Criteria 

The “reactor core cooling” criterion or “spent fuel pool cooling” criterion in 

10 CFR 50.150(a)(1) is satisfied if design features have been included in the design 

of the plant to specifically perform that cooling function with reduced use of operator 

action. 

Containment Criteria 

The “intact containment” criterion in 10 CFR 50.150(a)(1) is satisfied if the containment: 

(1) will not be perforated by the impact of a large, commercial aircraft; and (2) will 

maintain ultimate pressure capability, given a core damage event until effective 

mitigation strategies can be implemented. Effective mitigation strategies are those that 

provide, for an indefinite period of time, sufficient cooling to the damaged core or 

containment to limit temperature and pressure challenges below the ultimate pressure 

capability of the containment as defined in Section 19 of the DCD, Revision 3. 

Spent Fuel Pool Integrity Criteria 

The “spent fuel pool integrity” criterion in 10 CFR 50.150(a)(1) is satisfied if the impact of 

a large, commercial aircraft on the spent fuel pool wall or support structures would not 

result in leakage through the spent fuel pool liner below the required minimum water 

level of the pool. 

Reduced Operator Action 

The NRC considers the use of operator actions to be reduced when (1) all necessary 

actions to control the nuclear facility can be performed in the control room or at an 

alternate station containing equipment specifically designed for control purposes; and 

(2) a reduced amount of active operator intervention, if any, is required to meet the 

acceptance criteria in 10 CFR 50.150(a)(1). Reductions in the use of operator action are 

measured relative to the actions required to address aircraft impact without the aircraft 

impact assessment rule in place (e.g., similar actions contained in operational programs 

in place at current operating reactor sites). 

19S.4 Technical Evaluation 

The staff’s review of the proposed COL License Information Items 19.9k, 19.9l, 19.9m, and 

19.9n in the initial DCD amendment determined that these items should be part of the ITAAC. 

The applicant’s revised DCD amendment withdrew these COL license information items and 
19S-6

included additional ITAAC. These changes are consistent with the staff’s recommendation, and 

therefore acceptable. 

NRC staff reviewed the applicant’s description of key design features and the description 

of how the key design features show that the design meets the acceptance criteria in 

10 CFR 50.150(a)(1). The staff’s evaluation follows: 

19S.4.1 Reasonably Formulated Assessment 

The applicant states in the application that a design-specific assessment was performed with 

specific assumptions in NEI 07-13 Revision 7. The staff found this information incomplete 

because it was not clear that the applicant followed NEI 07-13 Revision 7 and NRC-endorsed 

guidance in whole or if there were any exceptions. The staff issued RAI 19-8 requesting the 

applicant to clarify whether the applicant fully followed NEI 07-13 or to state any exceptions. 

The applicant’s response dated February 18, 2010 (U7-C-STP-NRC-100040, ML100550027), 

clarifies that the applicant followed NEI 07-13 with no exceptions. The analysis was performed 

by the applicant’s subcontractor who has designed a nuclear power reactor facility either 

already licensed or certified by the NRC or currently under review by the NRC. Therefore, the 

staff found that the applicant has performed the assessment with a method that is reasonably 

formulated because the assessment was performed by a qualified organization, and it followed 

NEI 07-13 with no exceptions, and RAI 19-8 is therefore resolved. The staff verified that the 

applicant in the DCD AIA Amendment dated May 12, 2010 (U7-C-STP-NRC-100098, 

ML101340548) has followed NEI 07-13 without exception. Therefore, RAI 19-8 is closed. 

19S.4.2 Key Design Features for Core Cooling 

The key design features listed in the application that perform a core cooling-related function 

include (1) safety-related design features that are designed specifically to perform the core 

cooling function during normal power operation and following design-basis events initiated 

during power operation, and (2) non-safety-related design features that are designed specifically 

to support the core cooling function following the impact of a large commercial aircraft. The 

safety-related design features include the ECCS described in DCD Tier 2, Section 6.3; the 

SRVs described in DCD Tier 2, Section 5.2.2; the suppression pool described in DCD Tier 2, 

Section 6.2.1; and the COPS described DCD Tier 2, Section 6.2.5. The non-safety-related 

design features is the AFI system described in the application (Section 9.5.14). The preferred 

method of core cooling following the impact of a large commercial aircraft is to utilize normal 

safety-related injection systems and suppression pool cooling systems. However, if these 

systems are unavailable due to damage, the AFI system is used to add water to the reactor 

vessel and maintain the water level above the top of the active fuel, and heat is removed from 

the reactor vessel by relieving steam to the suppression pool via the SRVs. Steam is released 

to the outside atmosphere via the COPS after pressure in the suppression pool reaches the 

rupture pressure of the COPS rupture discs. 

NRC staff considered the descriptions of the safety-related features as well as staff‘s review of 

the ability of these features to perform their design-basis safety functions documented in the 

FSER supporting certification of the ABWR design (NUREG–1503). During the review, the staff 

confirmed that all of these design features can be initiated and operated from the control room 

or at an alternate station containing equipment specifically designed for control purposes and 

require little, if any, additional operator intervention to maintain the core cooling function. Based 

on this information, the staff found that the applicant has adequately described the use of these 
19S-7

systems to maintain core cooling following an impact from a large commercial aircraft when the 

reactor is shut down before the impact. 

In the initial review of the applicant’s descriptions, the staff noted that there was no description 

of design features or functional capabilities relied on to ensure that the acceptance criteria in 

10 CFR 50.150(a)(1) are met, if the reactor fails to shut down following either an automatic or 

manual actuation of the reactor protection system. The staff issued RAI 19-3 requesting the 

applicant to either describe how those design features and/or functional capabilities that will be 

credited for core cooling are able to be successful if the reactor fails to shutdown, or identify 

design features that ensure that the reactor will be shut down following the impact of a large, 

commercial aircraft. The applicant’s response to RAI 19-3 dated February 18, 2010 

(U7-C-STP-NRC-100043, ML100550025), states that the design of the facility locates the 

hydraulic control units (HCU) for reactor scram outside of the damage footprint that could be 

caused by the impact of a large commercial aircraft; and, as a result, advance warning to 

successfully shut down the reactor manually before impact may be assumed in accordance 

with the guidance in NEI 07-13. NEI 07-13 also requires that an assessment be made to 

determine if the physical damage from an aircraft impact could prevent the reactor from being 

shutdown in cases where advanced warning is not available. In response to RAI 15.08-1 dated 

February 18, 2010 (ML100550027) the applicant states that ABWR has two divisions of HCUs 

and both are located below grade and are not subject to physical damage from an aircraft 

impact. The staff reviewed the description of the HCUs in DCD Tier 2, Revision 3 and found 

that the HCUs are located below ground level and outside of the structural damage footprint and 

are designed to fail in a safe manner, when subjected to fire (i.e., the loss of function results in 

reactor scram). The staff also reviewed NEI 07-13, Revision 7 and confirmed that the guidance 

indicates that an assumption of a successful scram may be made when the equipment required 

is not expected to be affected by the damage caused from an impact of a large commercial 

aircraft. The applicant’s response indicates that the application will be modified to reflect this 

response. Therefore, based on the review described above, the staff found the applicant’s 

response acceptable because it adequately describes how the design features will ensure 

the AIA criteria are met. The staff verified that the revised DCD AIA Amendment dated 

May 12, 2010 (ML101340548) states that the design of the facility identifies the hydraulic control 

units (HCU) for the reactor scram as outside of the damage footprint that could be caused by 

the impact of a large commercial aircraft. As a result, the operator would have advance warning 

to successfully shut down the reactor manually before impact may be assumed in accordance 

with the guidance in NEI 07-13. Therefore, RAI 19-3 is closed. 

As discussed in the application, the AFI is designed to inject unheated water into the reactor 

vessel at high pressure (i.e., lowest opening pressure of any SRV) at a flow rate of 800 gpm. 

Considering the decay power of the ABWR, the staff found that the designated flow rate will be 

adequate to make up for a loss of inventory due to steam production from decay heat. 

The applicant did not describe in the DCD the method for removing heat from the reactor vessel 

when the AFI system is being used to inject water. Consequently, the staff issued RAI 19-2 

requesting the applicant to describe how heat is removed from the reactor vessel when the 

ECCS are unavailable and the AFI is being used to inject water into the reactor vessel. The 

applicant’s response to RAI 19-2 dated February 18, 2010 (U7-C-STP-NRC-100043, 

ML100550025), states that SRV cycling at high pressure will relieve energy from the reactor 

vessel to the suppression pool, and as the suppression pool heats up, pressure in the 

containment will increase until the COPS rupture disc opens, which will relieve steam to the 

atmosphere. The staff found this response describes a viable heat removal method because it 
19S-8

utilizes safety-related systems and components specifically designed to perform the functions 

needed to take steam from the reactor vessel and vent it to the outside atmosphere. 

The applicant’s response to RAI 19-2 also proposes a modification to the application that adds a 

more detailed description of the equipment used for decay heat removal and the heat removal 

pathways. The staff reviewed the proposed descriptions and confirmed that they completely 

describe the methods of heat removal and plant equipment relied upon to implement the 

methods. Therefore, the staff found the proposed modification acceptable. The staff verified 

that the applicant has included complete descriptions of the methods of heat removal and plant 

equipment relied upon to implement the methods in the revised application dated May 12, 2010 

(ML101340548). The staff confirmed that the final DCD AIA Amendment provides the proposed 

DCD changes. Therefore, RAI 19-2 is resolved and closed. 

The application states that the water source for the AFI will have a minimum capacity of 

300,000 gallons of water. The staff issued RAI 19-7 requesting the applicant to provide the 

basis for choosing this capacity and if this capacity was large enough to support core cooling 

until mitigation measures can be implemented that provide long-term cooling. The applicant’s 

response to RAI 19-7 dated February 18, 2010 (U7-C-STP-NRC-100043, ML100550025), 

states that the capacity was sized to provide sufficient make-up water to offset the calculated 

boil-off of water for 24 hours following scram from 100 percent power, assuming that the AFI is 

the only system available to provide core cooling. The applicant adds that 24 hours will be 

sufficient to replenish the AFI water source or initiate a firewater addition system for long-term 

cooling. The staff considers 24 hours to be a sufficient amount of time to implement mitigation 

measures for long-term core cooling for the following reason. Separate and apart from the 

requirements in 10 CFR 50.150, power reactor licensees are required, per 10 CFR 50.54(hh)(2), 

to develop and implement strategies for maintaining core cooling under conditions associated 

with the loss of large areas of the plant due to explosion or fire. The staff expects that licensee 

compliance with these requirements will result in implementation of mitigation measures within 

24 hours. Therefore, the staff determined the applicant’s technical basis for the capacity of the 

AFI water source acceptable. Therefore, RAI 19-7 is resolved and closed. 

The staff issued RAI 19-5 requesting that the applicant to describe any support systems or 

components that the AFI system relies upon for successful operation, such as systems for pump 

cooling or lubrication. The purpose of this request is to ensure that the applicant has identified 

as key design features all pieces of equipment needed for core cooling. The applicant’s 

response to RAI 19-5 dated February 18, 2010 (U7-C-STP-NRC-100043, ML100550025), 

states that the AFI system does not rely on any special or unique auxiliary support systems or 

equipment. The applicant explains that the AFI pump relies only on the water it is pumping for 

its own cooling and does not use auxiliary cooling equipment. The applicant adds that an oil 

pumping system is not used to lubricate the AFI pump. Lubrication is performed manually as 

part of the standard maintenance program for the pump. Based on the applicant’s response, 

the staff found that with the exception of the electric power support system described in the 

application, there are no additional systems or components associated with the AFI system that 

need to be identified and described. The staff confirmed that the final DCD AIA Amendment 

provides the proposed DCD changes. Therefore, RAI 19-5 is resolved and closed. 

Based on the description in the application, the AFI system includes dedicated I&Cs to allow 

operators to start and control the system and to monitor the pressure and water level in the 

reactor vessel to ensure that the system is performing properly. The staff found that the 

inclusion of these features in the design allows the AFI to be used to maintain core cooling with 

reduced operator action. 
19S-9

Based on the information in the application and the responses to the above RAIs, the staff found 

that the applicant has adequately described how core cooling will be maintained for at power 

scenarios using the AFI system in conjunction with the SRVs, the suppression pool, and the 

COPS. 

In the initial review of the applicant’s descriptions, the staff noted that the applicant did not 

include a description of the design features or functional capabilities relied upon to ensure that 

the acceptance criteria in 10 CFR 50.150(a)(1) are met, while the plant is shut down and the 

reactor core is being cooled via the shutdown cooling system with a large vent2

 in the primary 

system. The staff issued RAI 19-1 requesting the applicant to describe those design 

features and/or functional capabilities relied upon to ensure that the acceptance criteria in 

10 CFR 50.150(a)(1) are met, while the plant is shut down and the reactor core is being 

cooled via the shutdown cooling system. The applicant’s response to RAI 19-1 dated 

February 18, 2010 (U7-C-STP-NRC-100043, ML100550025), proposes a modification to the 

application stating that the existing RHR system will provide sufficient decay heat removal, but if 

the existing RHR system is unavailable as a result of the damage cause by the impact from a 

large commercial aircraft, decay heat removal will be accomplished by evaporation, with 

makeup provided by the AFI system. The staff considered the use of the RHR and AFI to be 

acceptable means of maintaining core cooling during shutdown conditions because those 

systems are designed to remove decay heat from the reactor and are capable of performing this 

function under plant conditions that exist with the a reactor shutdown for refueling. The staff 

verified that the revised DCD AIA Amendment dated May 12, 2010 (ML101340548) states that

the existing RHR system will provide sufficient decay heat removal unless the existing RHR 

system is unavailable as a result of the damage cause by the impact from a large commercial 

aircraft, in which case, decay heat removal will be accomplished by evaporation with makeup 

provided by the AFI system. The staff confirmed that the final DCD AIA Amendment provides 

the proposed DCD changes. Therefore, RAI 19-1 is resolved and closed. 

The applicant’s description of the instrumentation included with the AFI system states that 

instrumentation exists to provide information to the operator in the AFI pump house for reactor 

vessel water level, reactor pressure, suppression pool water level, and wetwell pressure. The 

staff found this description acceptable because this is the minimum information needed to 

maintain core cooling by an operator during water injecting into the reactor vessel. 

The applicant’s description of the AFI system power supply states that it is designed to power 

the AFI pump and motor operated valves and will meet non-Class 1E design requirements for 

non-safety-related power supplies. The staff found this description acceptable because the 

pump and valves are the only components of the system requiring motive power and because 

key design features included to satisfy the requirements of 10 CFR 50.150 are not required to 

be safety related. 

The staff considered the potential for re-criticality of the reactor core and thermal shock to the 

reactor vessel caused by cold water injected with the AFI system. The staff compared the 

description of the AFI system with the description of the RCIC system. The RCIC system is part 

of the certified ABWR design and designed to inject water as cold as 10 ̊C from a storage tank 

into the reactor vessel when the reactor coolant system is at pressures above normal operating 

pressure. In addition, the Commission has determined that the certified ABWR design, 

 

2

 A large vent is one of sufficient size to ensure that gas, including steam, can be removed from the 

reactor coolant system and water can be added to it should a loss of decay heat removal or loss 

of inventory event occur when the system is in cold shutdown.
19S-10

including the RCIC system, satisfies its requirements for emergency core cooling and fracture 

toughness. The staff found that the RCIC and the AFI systems are very similar in functional 

design. Therefore, it is reasonable to conclude that the AFI system can perform its core cooling 

function successfully without complications due to re-criticality or thermal shock. 

During the initial review of the amendment the staff determined that the description of the AFI 

water source was not adequate. The staff issued RAI 19-4 requesting the applicant to provide 

additional descriptive information regarding the design of the water source, and request that this 

information be included in the amended DCD. In the RAI the staff stated that if the applicant 

intended to only specify the functional capability of the water source and leave it to the COL 

applicant referencing the DC to provide this information as part of its detailed design, then the 

applicant for design certification should propose an ITAAC specifying the acceptance criteria 

that must be met to ensure compliance with 10 CFR 50.150. Inclusion of the ITAAC would 

require the COL Holder to ensure that the design of the water source meets the acceptance 

criteria before operation. In the response to this RAI dated February 18, 2010 (ML100550025), 

the applicant indicates that it will provide an ITAAC for the AFI water source regarding the 

capacity of the water source and its location relative to structures in the power block. The 

applicant provides a proposed ITAAC for the AFI water source in the response to RAI 14.02-1 

dated April 8, 2010 (ML101040254). The staff’s review of the response to RAI 14.02-1, 

including the proposed ITAAC to be included in the DCD Tier 1, is documented in 

Section 19S.4.5 of this SER. The staff found the applicant’s response to RAI 19-4 acceptable 

because the applicant has provided a proposed ITAAC that addresses the capacity of the water 

source and its location relative to structures in the power block. The staff reviewed the revised 

DCD AIA Amendment (ML101340548) and confirmed that the proposed changes are 

incorporated. Therefore, RAI 19-4 is resolved and closed. 

The staff also determined that the description of the AFI power source was not adequate. The 

staff issued RAI 19-6 requesting the applicant to propose an ITAAC that addresses the location 

of the power source relative to structures in the power block and identify any time delay in 

providing power necessary to operate the AFI system. In the response to this RAI dated 

February 18, 2010 (ML100550025), the applicant indicated that it would provide an ITAAC for 

the AFI power source regarding the location of the power source relative to structures in the 

power block. The applicant provided a proposed ITAAC for the AFI power source in the 

response to RAI 14.02-1 (ML101040254). The staff’s review of the response to RAI 14.02-1, 

including the proposed ITAAC to be included in the DCD Tier 1, is documented in 

Section 19S.4.5 of this SER. In the response, the applicant indicates that the time delay for 

providing power to the AFI was not addressed in the ITAAC because the design specification 

that injection with AFI be initiated within 30 minutes of an aircraft impact- described in 

Subsection 9.5.14.1 of the amendment– accounts for any delay in delivering motive power to 

the pump and valves in the AFI system. The staff found this response to be an acceptable 

rationale for not addressing the time delay for providing power to the AFI in the ITAAC because 

the time delay is captured in overall time delay which is described explicitly in the AFI design 

description which can be incorporated by reference in a COL application. In addition, the staff 

found the applicant’s response to RAI 19-6 acceptable because the applicant has provided a 

proposed ITAAC that addresses the location of the power source relative to structures in the 

power block and an explanation for not addressing the power system time delay in the ITAAC 

which the staff finds acceptable, as discussed above. The staff reviewed the revised DCD AIA 

Amendment (ML101340548) and confirmed that the proposed changes are incorporated. 

Therefore, RAI 19-6 is resolved and closed. 
19S-11

Furthermore, the staff issued RAI 19-11 requesting the applicant to provide descriptive 

information and clarification on several statements in DCD Section 19S.4.2 regarding the AFI 

power source, the separation between the AFI power source, and the emergency power 

sources in the power block. In the response to this RAI dated February 25, 2010 

(ML100600410), the applicant provides adequate clarification and indicates that it will provide 

an ITAAC for the AFI power source regarding the location of the power source relative to 

structures in the power block. The applicant provides a proposed ITAAC for the AFI power 

source in the response to RAI 14.02-1. The staff’s review of the response to RAI 14.02-1, 

including the proposed ITAAC to be included in the DCD Tier 1, is documented in 

Section 19S.4.5 of this SER. The staff found the applicant’s response to RAI 19-11 acceptable 

because the applicant has clarified statements in the amendment as requested by the NRC and 

provided a proposed ITAAC that addresses the location of the AFI power source relative to 

structures in the power block, including the normal emergency power sources. The staff 

reviewed the revised DCD AIA Amendment (ML101340548) and confirmed that the proposed 

changes are incorporated. Therefore, RAI 19-11 is resolved and closed. 

19S.4.3 Key Design Features that Protect Core Cooling Design Features 

Fire Protection 

The key design features of the Fire Protection Program that protect the core cooling key design 

features include all of the 3-hour, fire-rated barriers and 5-psid barriers in the reactor building 

and the control building, as described in DCD Tier 2, Revision 3, Sections 9.5.1 and 9A.4, 

Appendix 19S and Figures 9A.4-4 through 9A.4-8. In the initial review of the descriptions 

provided by the applicant, the staff noted that the applicant did not include the overpressure 

ratings for each credited fire door. NRC staff issued RAI 19-12 requesting the applicant to 

identify, in the DCD, the overpressure ratings for each credited fire door. The applicant’s 

response to RAI 19-12 in a letter dated February 18, 2010 (U7-C-STP-NRC-100040, 

ML100550027), identifies those doors that are identified as watertight doors, as shown in DCD 

Tier 2, Revision 3, Figure 1.2-8, which will be rated for at least 5 psid. In addition, the applicant 

clarifies that all other doors will not be rated at the minimum 5 psid. The staff considers 

watertight doors to have an equivalent rating of at least 5 psid, which provides an adequate 

means of confining deflagration overpressures as described in NEI 07-13 and an acceptable 

overpressure description. All watertight doors must maintain a fire rating that is consistent with 

the fire barrier where they reside. 

In the initial review of the descriptions provided by the applicant, the staff noted that the 

applicant did not identify the reactor building R/B floor plugs as key design features. These floor 

plugs separate the ground floor from the lower elevations that contain ECCS equipment. The 

floor plugs were added to limit the fire spread and protect key core cooling features. The staff 

issued RAI 19-13 requesting the applicant to identify these floor plugs as key design features in 

the DCD. The applicant’s response to RAI 19-13 in a letter dated February 18, 2010 

(U7-C-STP-NRC-100043, ML100550025), states that it will include these floor plugs as key 

design features and to identify them as such in the DCD. The staff considered the applicant’s 

responses acceptable. In the responses to both RAI 19-12 and RAI 19-13, the applicant 

provided a proposed DCD markup that the staff found acceptable. The staff confirmed that the 

revised DCD AIA Amendment provides the proposed DCD changes. Therefore, RAI 19-12 and 

19-13 are resolved and closed. 

During the review of DCD amendment submittal dated June 30, 2009 (ML092040048), NRC 

staff discovered an issue related to the NEI 07-13, Revision 7, “Fire Spread Rule” set. 
19S-12

Specifically, the applicant specified only the pressure rating of fire doors within the barriers that 

the applicant was crediting for withstanding overpressure effects. The staff’s position is that the 

entire barrier needs to be rated at least 5-psid when taking credit for a single barrier. This 

includes fire doors, penetration seals, hatches, etc. The basis for the staff’s position is that an 

overpressure may damage the components of the credited 3-hour fire-rated barrier, allowing fire 

to propagate beyond the credited barrier and beyond the fire damage footprint. In addition, 

proper identification and adequate description of these design features are required by 

10 CFR 50.150. Therefore, the staff issued RAI 19-14 requesting the applicant to state within 

the DCD the pressure rating of all barriers and barrier components identified under the one 

barrier option of the NEI 07-13 Fire Spread Rule set. The applicant’s response to RAI 19-14 

dated June 21, 2010 (U7-C-STP-NRC-100152, ML101750070), states that Subsection 9.5.1.1.3 

will be updated to state that the structural walls, floor, ceilings, and penetration seals and 

hatches within all fire barriers of the reactor building will be rated to withstand a 5 psid pressure 

differential. In the response to RAI 19-14, the applicant referenced a proposed DCD markup in 

a letter dated June 17, 2010 (U7-C-STP-NRC-100139, ML101720306) and provided a 

supplemental markup that the staff found acceptable. The proposed DCD includes the 

structural floor and ceilings as well as hatches in its discussion regarding the capability of 

certain fire protection features to withstand a 5 psid pressure differential. The staff confirmed 

that the final DCD AIA Amendment provides the proposed DCD changes. Therefore, RAI 19-14 

is resolved and closed. 

In the DCD amendment submittal dated June 30, 2009 (ML092040048), the applicant did not 

describe all fire doors adequately. The applicant described some doors as rated for 5-psid 

overpressure. These doors should also be rated as 3-hour fire-rated doors to meet the existing 

Fire Protection Program and fire hazards analysis requirements, as well as the guidance in 

NEI 07-13. In addition, proper identification and adequate description of these design features 

are required by 10 CFR 50.150. Therefore, NRC staff issued RAI 19-15 requesting the 

applicant to revise the DCD to indicate a 3-hour fire-rating for those 5psid doors identified under 

NEI 07-13. The staff and the applicant have had discussions dating back to January 27, 2010 in 

which the following DCD statement was discussed “...either a 5-psid door or two 3-hour rated 

fire doors...” The applicant should decide which option to be certified within the DCD. 

Proper identification and adequate description of these design features are required by 

10 CFR 50.150. For example, Room 512 within the DCD shows access from corridor A via a 

vestibule. The applicant appeared to be replacing the non-rated fire door with a 3-hour firerated door without describing whether the appropriate vestibule walls also have 3-hour 

fire-rating. Therefore, the staff requested the applicant to revise the DCD by choosing one 

option for each situation. The applicant’s response to RAI 19-15 dated June 21, 2010 

(U7-C-STP-NRC-100152, ML101750070), states that Appendix 9A will be updated to include 

the 3-hour fire rating to all 5 psid doors required to be fire doors. In addition, the applicant will 

remove all “...either a 5-psid door or two 3-hour rated fire doors” statements within 

Appendix 9A. Each removed statement will be replaced with the actual design feature utilized 

within the applicant’s assessment. In the response to RAI 19-15, the applicant referenced a 

proposed DCD markup in a letter dated June 17, 2010 (U7-C-STP-NRC-100139, 

ML101720306) that the staff found acceptable. The staff confirmed that the final DCD AIA 

Amendment provides the proposed DCD changes. Therefore, RAI 19-15 is resolved and 

closed. 

Furthermore, the applicant included in the revised submittal dated May 12, 2010 

(ML101340548) that the AFI [instrumentation] cabling itself will be 3-hour fire rated as it 

traverses through the reactor building and out to the AFI pump house. The staff accepts fire 

rated cabling, as opposed to cable wrapping, as this will ensure any AFI [instrumentation] 
19S-13

cabling exposed to a fire-induced overpressure situation will still maintain the integrity of the fire 

protection rating. In the applicant’s submittal of the amended ABWR DCD Revision 3, dated 

September 23, 2010 (ML102870017), the applicant added that the AFI instrumentation lines will 

be routed through rooms protected from fire exposure with the exception of Room 230. In this 

room, STPNOC states the AFI instrumentation line will be protected from fire utilizing a 3-hour 

fire rated wrap. The staff finds the routing of these instrumentation lines and limited use of fire 

rated wrap acceptable as the routing will ensure lines are not exposed to fire and the wrap is 

acceptable. This use of fire rated cable wrap is acceptable in this case because this room is not 

exposed to any overpressure but rather fire at ambient pressure only. The staff accepts that no 

other fire protection is required for the AFI system since no other portions of the AFI system 

which are vulnerable to fire reside within the reactor or control buildings. 

These fire protection features protect the credited core cooling equipment described in 19S.4.2 

above and protect the AFI instrumentation lines and cabling located in the reactor building. 

Based on the information in the application and the responses to the above RAIs, the staff found 

that the applicant has adequately described the fire protection key design features for 

maintaining core cooling in Sections 9.5.1, 9A and 19S. 

Primary Containment Structure 

In DCD Tier 2, Revision 3, Appendix 19S, the applicant states that the primary containment is a 

key design feature that will provide physical protection to the safety systems located inside the 

primary containment. NRC staff reviewed the information in DCD Tier 2, Revision 3 general 

arrangement drawings (Figure 1.2-1); containment and reactor building drawings (Figures 1.2-2, 

1.2-2a, 1.2-8 through 1.2-12, 2.15.10j, and 2.15.10k); DCD Tier 2; and Sections 3.8 and 3H.1. 

The applicant states that the primary containment is entirely surrounded by the reactor building 

structure and therefore, a direct impact on the primary containment from a large commercial 

aircraft is not possible. The staff verified the applicant’s statement by reviewing the cited 

drawings. The applicant determined by analysis that a strike on the primary containment 

(1) would not result in the perforation of the primary containment, and (2) would not cause direct 

damage to the systems within the containment or expose them to jet fuel. 

Based on this review, the staff concluded that the applicant has adequately described the 

primary containment as a key design feature for protecting safety systems inside the primary 

containment to maintain core cooling. 

Reactor Building Structure 

In DCD Tier 2, Revision 3, Appendix 19S, the applicant states that the location and design of 

the reactor building structure are key design features that protect portions of the primary 

containment and the south wall of the control building from the impact of a large, commercial 

aircraft. This includes the protection provided by exterior walls, interior walls, intervening 

structures and barriers on the large openings in the reactor building exterior walls. The 

applicant states that ABWR containment is integrated with, and fully contained within, the 

reactor building. The containment and the reactor building are supported by a 5.5m thick 

common foundation mat. The bottom of the foundation mat is embedded in the ground 

approximately 26m below grade. Figure 1.2-1 shows the location of the reactor building in 

relation to other plant structures. The staff reviewed the information in DCD Tier 2, Revision 3 

general arrangement drawings (Figures 1.2-1 1.2-2, 1.2-2a, 1.2-8 through 1.2-12 2.15.10j 

and 2.15.10k ), Section 3.8, and 3H.1, and verified the accuracy of the above description of the 

reactor building. 
19S-14

Based on the above review, the staff found adequate (in level of detail and scope) and 

acceptable the applicant’s description of the reactor building and its design, as a key design 

feature for protecting portions of the primary containment and the south wall of the control 

building from the impact of a large, commercial aircraft. 

Turbine Building 

In DCD Tier 2, Revision 3, Appendix 19S, the applicant states that the location and design of 

the turbine building structure, as described in DCD Tier 1, Section 2.15.11 and shown in DCD 

Tier 2, Revision 3, general arrangement drawings Figure 1.2-1, and Figures 1.2-24 through 

1.2-31 are key design features that protect portions of the north wall of the control and reactor 

building from the impact of a large, commercial aircraft. The turbine building is a Non-seismic 

Category I structure located to the north of the control building. The staff reviewed DCD Tier 2, 

Revision 3, general arrangement drawings, the above noted figures, Section 3.8.4 and 

Section 3H.2. The review led the staff to conclude that the above referenced drawings and 

documents contain acceptable level of detail and information to adequately describe the turbine 

building structure. Based on its review the staff found adequate and acceptable the applicant’s 

description of the turbine building as a key design feature for protecting portions of the north 

wall of the control and reactor building from the impact of a large, commercial aircraft. 

Control Building Structures 

In DCD Tier 2, Revision 3, Appendix 19S, the applicant states that the location and design of 

the control building structure as shown in DCD Tier 2, Revision 3, (general arrangement 

drawings, Figure 1.2-1); Section 3.8.4; and Section 3H.2 are key design features that protect the 

reactor building from the impact of a large commercial aircraft. NRC staff reviewed the above 

information and control building-related figures to confirm the applicant’s statement. Based on 

this review, the staff found adequate and acceptable the applicant’s description of the control 

building as a key design feature for protecting portions of the north wall of the reactor building 

from the impact of a large commercial aircraft. 

The staff found that the applicant has adequately described the key design features for 

providing physical protection to the reactor building by maintaining core cooling. 

19S.4.4 Key Design Features that Maintain Integrity of the Spent Fuel Pool 

The key design features credited to maintain the integrity of the spent fuel pool (SFP) are the 

location and design of the SFP and its support structure, as described in DCD Tier 2, 

Revision 3, Sections 1.2 and 9.1 and Figure 1.2-12. The applicant indicates that the location 

and design of the SFP and its support structure ensure that the SFP can withstand the effects of 

an impact of a large commercial aircraft. In the initial review of the DCD amendment (Tier 2, 

Appendix 19S), NRC staff noted that the applicant provided no statement supporting a 

conclusion that there would be no leakage from the SFP liner that would allow the SFP to drain 

below the required minimum water level, in accordance with the sufficiency criteria in NEI 07-13. 

The staff issued RAI 19-9 requesting the applicant to identify any scenarios resulting in SFP 

liner leakage and drainage below the required minimum water level. The applicant’s response 

to RAI 19-9 in a letter dated February 18, 2010 (U7-C-STP-NRC-100040, ML100550027), 

clarifies that there are no impact scenarios that will result in leakage from the SFP liner below 

the required minimum water level. The staff issued RAI 19-10 requesting the applicant to 

describe, in the DCD, how the SFP location and its design protect the integrity of the SFP and 

prevent perforation below the required minimum water level. The applicant’s response to 
19S-15

RAI 19-10 in a letter dated February 18, 2010 (ML100550027), clarifies that the SFP is located 

entirely within the reactor building and a detailed analysis of an aircraft impact showed that the 

structural design—which includes the SFP walls, liner, and support structures—is adequate to 

prevent a rupture of the liner. In addition, the applicant states that all pipes are configured so 

that they will not allow drainage below the minimum water level. The staff found that this 

clarification will allow the integrity of the SFP to be maintained. In the responses to both 

RAI 19-9 and RAI 19-10, the applicant provides a proposed DCD markup that the staff found 

acceptable. The staff confirmed that the revised DCD AIA Amendment (ML101340548) 

provides the proposed DCD changes; therefore, RAI 19-9 and RAI 19-10 are resolved. The 

staff found that the applicant has adequately described the key design features for ensuring the 

integrity of the SFP. The staff determined that the description of the key design features is in 

compliance with 10 CFR 50.150 and is therefore adequate and acceptable. Therefore, RAI 19-9 

and RAI 19-10 are closed. 

19S.4.5 Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) 

During the initial review of DCD amendment submittal dated June 30, 2009 (ML092040048), 

NRC staff discovered the applicant failed to provide sufficient ITAAC to verify that the AFI 

system’s key design characteristics and performance requirements are verified to ensure that 

the AFI system and its supporting systems will be available when required following an aircraft 

impact event. 

The staff issued RAI 14.02-1 requesting the applicant to propose ITAAC for the AFI system that 

are necessary and sufficient to provide reasonable assurance that, if the inspections, tests, and 

analyses are performed and the acceptance criteria are met, a facility referencing the amended 

and certified ABWR design is constructed and will be operated in conformity with the design 

certification and NRC regulations. The applicant’s response to RAI 14.02-1 dated April 8, 2010 

(U7-C-STP-NRC-100049, ML101040254) provides several ITAAC for the AFI system in DCD 

Tier 1, Section 2.11.24 and Table 2.11.24: 

ITAAC# 2.11.24-1 states that the basic configuration of the AFI system is as described in 

Section 2.11.24. The staff found this response acceptable because Section 2.11.24 

provides the high-level objective and design requirements of this system that the COL 

applicant must adhere to. 

ITAAC# 2.11.24-2 states that the AFI pump is capable of injecting ≥800 gpm into the RPV at 

the lowest SRV safety lift pressure. The staff found a minimum flow rate of 800 gpm to be 

an acceptable flow rate in order to maintain core cooling. The staff found the injection 

pressure of the lowest SRV safety lift pressure to be acceptable because that is the 

maximum pressure the AFI system is reasonably anticipated to be used for. 

ITAAC# 2.11.24-3 states that the AFI system water supply has a minimum capacity of 

300,000 gallons and is refillable. This amount of water (300,000 gallons) is expected to 

provide 24 hours of core cooling capability, as discussed in Section 19.S.4.2 of DCD, which 

the staff found acceptable. A refillable water supply does not hinder but rather enhances the 

safety capabilities of the AFI system and therefore is an acceptable acceptance criterion. 

ITAAC# 2.11.24-4 through 2.11.24-6 state that the AFI pump house, AFI water supply, and 

AFI power supply (respectively) are located a minimum of 300 feet from the nearest outside 

wall of each building: the reactor building, control building, and turbine building. The staff 

found this description acceptable because the AFI pump house, which houses most of the 
19S-16

AFI equipment, together with the water supply and power supply are required to survive 

certain impact scenarios that render the ECCS inoperable. This separation distance is 

necessary so that both the target area and AFI pump house are not damaged by a single 

post-impact explosion. At least one of these systems is required in order to maintain core 

cooling. 

ITAAC# 2.11.24-7 through 2.11.24-9 state that barriers defined in NEI 07-13, Revision 7 

exist between each AFI pump house, AFI water supply, and AFI power supply (respectively) 

and between each building: the reactor building, control building, and turbine building. The 

staff found this description acceptable because the AFI pump house, AFI water supply, and 

AFI power supply, together, are required to survive certain impact scenarios where the 

reactor building, control building, and/or turbine building have not. These intervening 

barriers defined in NEI 07-13, Revision 7 ensure that the physical impact of an aircraft does 

not destroy both the AFI system and the ECCS. At least one of these systems is required to 

survive in order to maintain core cooling. 

ITAAC# 2.11.24-10 states that instrumentation exists to provide information to the operator 

in the AFI pump house for reactor vessel water level, reactor pressure, suppression pool 

water level, and wetwell pressure. The staff found this description acceptable because this 

is the minimum information needed by an operator injecting water into the core in order to 

maintain core cooling. 

ITAAC# 2.11.24-11 states that MOVs in the AFI system injection line operate as designed 

on a manual initiation signal. The staff found this description acceptable because the AFI 

system is a manually operated system and therefore these MOVs, which are normally 

closed, will need to be opened upon the AFI operator’s request for the system to be 

successful. There are no automatic signals required because this system is used solely for 

the purpose of aircraft impact events which the NRC classifies as a beyond-design-basis 

event. 

The staff reviewed the information in the revised DCD Tier 1, Section 2.11.24 and Table 2.11.24 

in accordance with the guidance in SRP Section 14.3. The staff found that the applicant’s 

response to RAI 14.02-1 and the proposed ITAAC, as shown in DCD Tier 1, Section 2.11.24 

and Table 2.11.24 meet the ITAAC requirements (10 CFR 52.47(b)(1)), and provide a 

necessary and sufficient basis for the staff to conclude that there is reasonable assurance that, 

if the inspections, tests, and analyses are performed and the acceptance criteria are met, a 

facility referencing the amended and certified ABWR design is constructed and will be operated 

in conformity with the design certification and applicable regulations, including 10 CFR 50.150. 

The staff also found that the aggregate of ITAAC representing the AFI system are 

comprehensive and encompass the necessary functions to ensure the AFI system is capable of 

performing its intended function following an aircraft impact event. Furthermore, the staff found 

that the ITAAC descriptions are objective, verifiable, and consistent with the Tier 2 information. 

Based on this review and a review of the selection methodology and criteria for development of 

the Tier 1, the staff concluded that the top-level design features and performance characteristics 

of the AFI system are appropriately described in Tier 1, Section 2.11.24 and Table 2.11.24 and 

this information is acceptable, therefore, the applicant’s response is acceptable. The staff 

verified that the revised DCD AIA Amendment (ML101340548) includes the proposed ITAACs, 

and RAI 14.02-1 is resolved and closed. 
19S-17

19S.5 Conclusion 

The NRC staff’s review confirmed that the applicant has performed an aircraft impact 

assessment that is reasonably formulated to identify design features and functional 

capabilities to show, with reduced use of operator action, that the acceptance criteria in 

10 CFR 50.150(a)(1) are met. The staff found that the applicant has adequately described the 

key design features credited to meet 10 CFR 50.150, including descriptions of how the key 

design features show that the acceptance criteria in 10 CFR 50.150(a)(1) are met. Therefore, 

the staff found that the applicant meets the applicable requirements of 10 CFR 50.150(b). 

20-1

20 GENERIC ISSUES 

20.5.1.3 Identification of Potential Design Improvements 

20.5.1.3.1 Introduction 

This section of the SER documents the evaluation of potential design improvements consistent 

with information provided in the ABWR DCD FSER (NUREG–1503). Neither the certified 

ABWR DCD, nor the AIA amendment, contains Section 20.5. 

On June 30, 2009, STPNOC submitted an application to amend the Design Certification 

Rule for the U.S. ABWR, (STPNOC Letter No. U7-C-STP-NRC-090070, ML 092040048) to 

address the requirements of 10 CFR 50.150, the Commission's new aircraft impact rule 

(10 CFR 50.150). As part of this application, in ABWR DCD Revision 3, STPNOC is proposing 

to add a new AFI system and pump house in addition to other equipment to meet the 

requirements of the new aircraft impact rule. On November 10, 2009, STPNOC submitted a 

supplemental environmental report, titled, "Applicant's Supplemental Environmental 

Report-Amendment to ABWR Standard Design Certification," to address the requirements of 

10 CFR 51.55(b), (STPNOC Letter No. U7-C-STP-NRC-090180, ADAMS Accession 

No. ML093170454). 

20.5.1.3.2 Summary of Application 

As part of the aircraft impact amendment that proposes to add a new AFI system and pump 

house in addition to other equipment, STPNOC submitted a supplemental environmental report 

to address the requirements of 10 CFR 51.55(b). In this report, the applicant evaluated the 

impacts of the design changes on the assessment of Severe Accident Mitigation Design 

Alternatives (SAMDAs). 

20.5.1.3.3 Regulatory Basis 

The regulatory basis for reviewing this section is 10 CFR 50.34(f). The information on the 

identification of potential cost-beneficial SAMDAs that supports this review was provided in the 

environmental report pursuant to 10 CFR 51.55(b). 

20.5.1.3.4 Technical Evaluation 

 NRC staff reviewed the applicant’s supplemental environmental report and checked the 

referenced certified ABWR DCD, and NUREG–1503, Supplement 1. 

The numerical values and discussion in the ABWR DCD FSER (NUREG-1503), 

Section 20.5.1.3 was based on the values in Section 19P of the Standard Safety Analysis 

Report (SSAR). As part of the design certification rulemaking, General Electric (GE) updated 

SSAR Section 19P, but did not include it in the DCD. Instead, GE relocated this discussion to 

GE's "Technical Support Document (TSD) for the ABWR,” Revision 1, December 1994, which 

was contained in an attachment to a letter from GE to the NRC dated December 21, 1994, 

(ADAMS Accession No. ML100210563). NRC staff reviewed the updated information in the 

TSD, and in FSER Supplement 1 Section 20.5.1.3, indicated that the conclusions in the final 

environmental assessment issued with the design certification rule will remain unchanged. 
20-2

The applicant’s environmental report evaluated the impacts of the DCD Revision 3 design 

changes on the assessment of SAMDAs, which were evaluated in the TSD. The applicant’s 

review concludes that the design changes do not result in identification of any new SAMDAs 

that could become cost beneficial. 

NRC staff reviewed the applicant’s evaluations of the design changes and confirmed that the 

new designs will not result in a change to the ABWR probabilistic risk assessment (PRA) or 

DCD Chapter 19. Therefore, the staff concluded that the new design would not impact the 

original PRA, which provides input to the SAMDA evaluation, and therefore, would not change 

the conclusions reached in NUREG–1503 Supplement 1 and in the final environmental 

assessment issued with the ABWR design certification rule. 

20.5.1.3.5 Conclusion 

Based on the above technical evaluations, NRC staff concluded that the applicant has 

adequately addressed 10 CFR 50.34(f). The staff determined the proposed design changes 

would not affect the conclusions reached in NUREG–1503 Supplement 1, Section 20.5.1.3. 

The staff determined that there would be no adverse impacts from complying with the 

requirements for consideration of aircraft impacts on conclusions reached by the NRC in its 

review of the original ABWR design certification (NUREG-1503). 
A-1

APPENDIX A 

REFERENCES 

American Standards of Mechanical Engineers (ASME): 

─ ─ ─ ─ ─, ASME NQA-1-1994, “Quality Assurance Requirements for Nuclear Facility 

Applications.” 

American National Standards Institute/American Nuclear Society (ANSI/ANS) 

— — — — —, ANSI/ANS 3.2-1999, “Administrative Controls and Quality Assurance for the 

Operational Phase of Nuclear Power Plants.” 

Institute of Electrical and Electronic Engineers (IEEE) Standards: 

─ ─ ─ ─ ─, IEEE Std 336-1985, “IEEE Standard Installation, Inspection, and Testing 

Requirements for Power, Instrumentation, and Control Equipment at Nuclear Facilities.” 

─ ─ ─ ─ ─, IEEE Std 498-1985, “IEEE Standard Requirements for the Calibration and Control of 

Measuring and Test Equipment Used in Nuclear Facilities.” 

─ ─ ─ ─ ─, IEEE Std 603-1991, “IEEE Standard Criteria for Safety Systems for Nuclear Power 

Generating Stations.” 

─ ─ ─ ─ ─, IEEE Std 603-1991, Clause 5.6.3, “Independence Between Safety Systems and 

Other Systems.” 

─ ─ ─ ─ ─, IEEE Std 603-1991, Clause 6.3, “Interaction Between the Sense and Command 

Features and Other Systems.” 

Nuclear Energy Institute (NEI) 

— — — — —, NEI 06-14, Revision 8, “Quality Assurance Program Description,” May 2010. 

— — — — —, NEI 07-13, Revision 7, “Methodology for Performing Aircraft Impact 

Assessments for New Plant Designs,” May 2009. 

U.S. Code of Federal Regulations (CFR): 

─ ─ ─ ─ ─, Title 10 CFR Part 21, “Reporting of Defects and Noncompliance.” 

─ ─ ─ ─ ─, Title 10 CFR Part 50, “Domestic Licensing of Production and Utilization Facilities.” 

─ ─ ─ ─ ─, Title 10 CFR Part 50, Appendix A, “General Design Criteria for Nuclear Power 

Plants.” 

─ ─ ─ ─ ─, Title 10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power 

Plants and Fuel Reprocessing Plants.” 

─ ─ ─ ─ ─, Title 10 CFR 50.34, “Contents of Applications; Technical Information.” 
A-2

─ ─ ─ ─ ─, Title 10 CFR 50.36, “Technical Specifications.” 

─ ─ ─ ─ ─, Title 10 CFR 50.48, “Fire Protection.” 

─ ─ ─ ─ ─, Title 10 CFR 50.55, “Protection and Safety Systems.” 

─ ─ ─ ─ ─, Title 10 CFR 50.70, “Inspections.” 

─ ─ ─ ─ ─, Title 10 CFR 50.71, “Maintenance of Records, Making of Reports.” 

─ ─ ─ ─ ─, Title 10 CFR 50.120, “Training and Qualification of Nuclear Power Plant Personnel.” 

─ ─ ─ ─ ─, Title 10 CFR 50.150, “Aircraft Impact Assessment.” 

─ ─ ─ ─ ─, Title 10 CFR 52.47, “Contents of Applications; Technical Information.” 

─ ─ ─ ─ ─, Title 10 CFR 52.63, “Finality of Standard Design Certifications.” 

─ ─ ─ ─ ─, Title 10 CFR 50.73, “Relationship to Other Subparts.” 

─ ─ ─ ─ ─, Title 10 CFR 52.79, “Contents of Applications; Technical Information in Final Safety 

Analysis Report.” 

─ ─ ─ ─ ─, Title 10 CFR Part 54, “Requirements for Renewal of Operating Licenses for Nuclear 

Power Plants.” 

─ ─ ─ ─ ─, Title 10 CFR Part 52, Appendix A, “Design Certification Rule for the U.S. Advanced 

Boiling-Water Reactor.” 

U.S. Nuclear Regulatory Commission (NRC) 

Generic Letter (GL) 

— — — — —, GL 85-06, “Quality Assurance Guidance for ATWS Equipment That Is Not 

Safety-Related,” April 16, 1985. 

— — — — —, GL 88-18, “Plant Record Storage on Optical Disks,” October 20, 1988. 

— — — — —, GL 89-02, “Actions to Improve the Detection of Counterfeit and Fraudulently 

Marketed Products,” March 21, 2009. 

— — — — —, GL 91-05, “Licensee Commercial-Grade Procurement and Dedication 

Programs,” April 9, 1991. 
A-3

NUREG-Series Reports: 

─ ─ ─ ─ ─, NUREG–0800 “Standard Review Plan for the Review of Safety Analysis Reports for 

Nuclear Power Plants,” June 1987. 

─ ─ ─ ─ ─, NUREG–1503, “Final Safety Evaluation Report Related to the Certification of the 

Advanced Boiling Water Reactor Design,” July 1994. 

─ ─ ─ ─ ─, NUREG–1503, Supplement 1, “Final Safety Evaluation Report Related to the 

Certification of the Advanced Boiling Water Reactor Design,” July 1997. 

Regulatory Guides: 

─ ─ ─ ─ ─, RG 1.8, Revision 3, “Qualification and Training of Personnel for Nuclear Power 

Plants,” May 2000. 

─ ─ ─ ─ ─, RG 1.26, Revision 4, “Quality Group Classifications and Standards for Water-, 

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants,” 

March 2007. 

─ ─ ─ ─ ─, RG 1.28, Revision 3, “Quality Assurance Program Requirements (Design and 

Construction),” August 1985. 

─ ─ ─ ─ ─, RG 1.29, Revision 4, “Seismic Design Classification,” March 2007. 

─ ─ ─ ─ ─, RG 1.33, Revision 2, “Quality Assurance Program Requirements,” February 1978. 

─ ─ ─ ─ ─, RG 1.37, Revision 1, “Quality Assurance Requirements for Cleaning of Fluid 

Systems and Associated Components of Water-Cooled Nuclear Power Plants,” March 2007. 

─ ─ ─ ─ ─, RG 1.68, Revision 3, “Initial Test Programs for Water-Cooled Nuclear Power 

Plants,” March 2007. 

─ ─ ─ ─ ─, RG 1.70, Revision 3, “Standard Format and Content of Safety Analysis Reports for 

Nuclear Power Plants (LWR Edition),” November, 1978. 

─ ─ ─ ─ ─, RG 1.155, “Station Blackout,” August 1988. 

─ ─ ─ ─ ─, RG 1.189, Revision 1, “Fire Protection for Nuclear Power Plants,” April 2001. 

─ ─ ─ ─ ─, RG 1.206, “Combined License Applications for Nuclear Power Plants,” July 2007. 

Draft Regulatory Guides: 

— — — — —, DG-1176 “Guidance for the Assessment of Beyond-Design-Basis Aircraft 

Impacts,” July 2009. 

 Regulatory Issue Summaries: 

— — — — —, RIS 2000-018, “Guidance on Managing Quality Assurance Records in Electronic 

Media,“ October 23, 2000. 
A-4

Atomic Energy Act of 1954, as Amended, Public Law 102-486 (106 Stat 2943), October 24, 

1992, Available at http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0980/

ADAMS Accession Numbers: 

ML100210563—Letter dated December 21, 1994, to Document Control Desk, Subject: 

NEPA/SAMDA Submittal for ABWR, with attachment titled, “Technical Support Document for 

the ABWR.” 

ML052360625—Memorandum dated August 26, 2005, from Dale, F. Thatcher, to L. Raghavan, 

Subject: Approval of Nuclear Management Company Change to Quality Assurance Report, 

NMC-1 (MC7585, MC7587, MC7588, MC7589, MC7590, MC7591, MC7592). 

ML003693241—Letter dated March 20, 2000, to Mr. H. L. Sumner, Jr., Vice President—Nuclear 

Hatch Project, Southern Nuclear Operating Company, Inc. Subject: Edwin I. Hatch Nuclear 

Power Station, Units 1 And 2, Re: Approval of Relief Request RR-27, Third-Year Interval 

Inservice Inspection Program (TAC Nos. MA6163 and MA6164). 

ML052710224—Letter dated September 28, 2005, to Mr. Gregg R. Overbeck, Senior Vice 

President, Nuclear Arizona Public Service Company. Subject: Palo Verde Nuclear Generating 

Station, Units 1, 2, and 3 - Approval of Change to Quality Assurance Program (CommercialGrade Calibration Services) (TAC Nos. MC4402, MC4403, and MC4404). 

ML070510300—Letter dated April 25, 2007, to Adrian P. Heymer, Senior Director, New Plant 

Deployment, Nuclear Generation Division, Nuclear Energy Institute. Subject: Final Safety 

Evaluation for Technical Report NEI 06-14, “Quality Assurance Program Description” (Project 

No. 689; TAC No. Md3406). 

ML073440472—Letter dated December 19, 2007, to Mr. Keith Greenaway, President/CEO, 

ACLASS Accreditation Services. Subject: Reply to Your Letter Dated September 26, 2007, 

Seeking Agency Assistance in Accepting ACLASS Accreditation Services. 

ML081140564—Letter dated April 22, 2008, to Mr. R. Douglas Leonard, Jr. Managing Director, 

Laboratory Accreditation Bureau. Subject: Reply to Your Letter Dated February 29, 2008, 

Seeking Assistance in Accepting Laboratory Accreditation Bureau. 

ML081330253—Letter dated May 14, 2008, to Mr. Patrick V. McCullen, Vice President, 

International Accreditation Service, Inc. Subject: Reply to Your Letter Dated March 3, 2008, 

Seeking Assistance in Accepting International Accreditation Service, Inc. 

ML092040048—Letter dated June 30, 2009, to Document Control Desk at NRC. Subject: 

South Texas Project Application to Amend the Design Certificate Rule for the U.S. Advanced 

Boiling-Water Reactor (ABWR). 

ML092650695—Letter dated November 3, 2009, to Mr. Russell J. Bell, Director New Plant 

Licensing, Nuclear Generation Division, Nuclear Energy Institute. Subject: Final Safety 

Evaluation for Technical Report NEI 06-14, “Quality Assurance Program Description,” 

Revision 7. 
A-5

ML093170454—Letter dated November 10, 2009, to Document Control Desk at NRC, Subject: 

South Texas Project, Units 3 & 4, Submittal of "Applicant’s Supplemental Environmental 

Report—Amendment to ABWR Standard Design Certification.” 

ML100190088—Letter dated January 13, 2010, to Document Control Desk at NRC. Subject: 

South Texas Project, Units 3 and 4 - Response to Request for Additional Information 

Question 06.02.04-1 Related to Application to Amend the ABWR DCD Amendment Part 2, 

Tier 2, Section 6.7 “High Pressure Nitrogen Gas Supply System” Provided in Attachment. 

ML100210563—Letter dated December 21, 1994, to Document Control Desk at NRC. Subject: 

“NEPA/SAMDA Submittal for the ABWR.” Letter from GE Nuclear Submitting Technical Support 

Document (TSD) for the ABWR. 

ML100250139—Letter dated January 20, 2010, to Document Control Desk at NRC. Subject: 

South Texas Project, Units 3 and 4, Response to Request for Additional Information Letter 

Nos. 3 and 4 Related to the Application to Amend the ABWR DCD Part 2, Tier 2, Sections 5.2 

and 7.7. 

ML100470589—Letter dated February 8, 2010, to the Document Control Desk at the NRC. 

Subject: South Texas Project, Unit 3 and 4, Response to Request for Additional Information 

(RAI) Letter Number 10 Related to the Application to Amend the ABWR DCD Part 2, Tier 2, 

Section 1.0 Provided in Attachment 1 to the Referenced Letter. 

ML100550025—Letter dated February 18, 2010, to Document Control Desk at NRC. Subject: 

South Texas Project, Unit 3 and 4, Transmittal of Response to Request for Additional 

Information (RAI) Letter Numbers 2 and 5 Related to the Application to Amend the ABWR DCD 

Part 2, Tier 2, Section 19, Provided in Attachment 1 to the Referenced Letter. 

ML100550027—Letter dated February 18, 2010, to Document Control Desk at NRC. Subject: 

South Texas Project, Unit 3 and 4, Transmittal of Response to Request for Additional 

Information (RAI) Letter Numbers 5, 6, 7, and 8, Related to the Application to Amend the ABWR 

DCD Part 2, Tier 2, Sections 3.8, 5.2, 15.8, and 19, Provided in Attachment 1 to the Referenced 

Letter. 

ML100600410—Letter dated February 25, 2010, to Document Control Desk at NRC. Subject: 

South Texas Project, Units 3 and 4, Response to Request for Additional Information RAI Letter 

Numbers 5, 8, and 9 Related to the Application to Amend the ABWR DCD Part 2, Tier 2, 

Sections 3.8, 9.2, and 19. 

ML100640162—Letter dated March 3, 2010, to Document Control Desk at NRC. Subject: 

South Texas Project, Units 3 and 4 - Response to Request for Additional Information. 

ML100770388—Letter dated March 17, 2010, to Document Control Desk at NRC. Subject: 

South Texas Project, Units 3 and 4 - Responses to Request for Additional Information. 

ML101040254—Letter dated April 8, 2010, to Document Control Desk at NRC. Subject: South 

Texas Project, Units 3 and 4—Responses to Request for Additional Information, Letter 

Number 11 Related to the Application to Amend the ABWR DCD Part 2, Tier 2, Section 14.2. 
A-6

ML101040345—Letter dated April 8, 2010, to Document Control Desk at NRC. Subject: South 

Texas Project, Units 3 and 4—Responses to Request for Additional Information, Related to the 

Application to Amend the ABWR DCD Part 2, Tier 2, Section 1.0 

ML101120085—Letter dated April 19, 2010, to Document Control Desk at NRC. Subject: 

South Texas Project, Units 3 and 4—Responses to Request for Additional Information 

Question 09.02.04-1. 

ML101190120—Letter dated April 26, 2010, to Document Control Desk at NRC. Subject: 

South Texas Project, Units 3 and 4 - Response to Request for Additional Information 

Question 03.02.02-1 Related to the Application to Amend the ABWR DCD Part 2, Tier 2, 

Section 3.2. 

ML101340548—Letter dated May 12, 2010, to Document Control Desk at NRC. Subject: South 

Texas Project Proposed ABWR DCD AIA Amendment Revision 1. 

ML101470298—Draft Letter from NRC to Mr. Scott M Head, South Texas Project Nuclear 

Operating Company, Subject: NRC Inspection Report Nos. 05200012/2010201 and 

5200013/2010201, dated July 21, 2010. 

ML101530610—Letter dated May 27, 2010, to Document Control Desk at NRC. Subject: South 

Texas Project Units 3 and 4—Responses to Request for Additional Information 

Questions 03.08.04-2 and 03.08.04-4 Related to the Application to Amend the ABWR DCD 

Part 2, Tier 2, Section 3.2. 

ML101720306—Letter dated June 17, 2010, to Document Control Desk at NRC. Subject: STP 

ABWR Aircraft Impact Assessment (AIA) Revision. 

ML101470298—Letter dated June 21, 2010, to Document Control Desk at NRC. Subject: 

South Texas Project Units 3 and 4—Responses to Request for Additional Information Letter 

Number 13 Related to the Application to Amend the ABWR DCD Part 2, Tier 2, Section 19. 

ML102000496—Letter dated July 12, 2010, to Document Control Desk at NRC. Subject: 

ABWR STP Aircraft Impact Assessment (AIA) Amendment Revision-0. 

ML102240435—Letter dated August 4, 2010, to Document Control Desk at NRC. Subject: 

ABWR STP Aircraft Impact Assessment (AIA) Amendment Revision-1. 

ML103190120—Letter dated September 2, 2010, to Document Control Desk at NRC. Subject: 

ABWR STP Aircraft Impact Assessment (AIA) Amendment Revision-2. 

ML102870017—Letter dated September 23, 2010, to Document Control Desk at NRC. Subject: 

ABWR STP Aircraft Impact Assessment (AIA) Amendment Revision-3. 
B-1

APPENDIX B 

CHRONOLOGY OF CORRESPONDENCE 

Document 

Date 

ADAMS 

Accession 

Number 

Subject Correspondence From To Docket 

Number 

6/30/2009 ML092040048 South Texas Project Application to 

Amend the Design Certificate Rule 

for the US Advanced Boiling Water 

Reactor (ABWR). 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

11/10/2009 ML093170454 South Texas Project, Units 3 & 4, 

Submittal of "Applicant's 

Supplemental Environmental ReportAmendment to ABWR Standard 

Design Certification." 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

11/10/2009 ML093170511 Transmittal Letter of 10/14/2009 

Meeting Presentation Materials. 

Letter 

Meeting Briefing 

Package/ 

Handouts 

Slides and 

Viewgraphs 

South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

1/13/2010 ML100190088 South Texas Project, Units 3 & 4 - 

Response to Request for Additional 

Information Question 06.02.04-1 

Related to Application to Amend the 

ABWR DCD Amendment Part 2, Tier 

2, Section 6.7 "High Pressure 

Nitrogen Gas Supply System" 

Provided in Attachment. 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

1/20/2010 ML100250139 South Texas Project, Units 3 and 4, 

Response to Request for Additional 

Information Letter Nos. 3 and 4 

Related to the Application to Amend 

the ABWR DCD Part 2, Tier 2, 

Sections 5.2 and 7.7. 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

2/8/2010 ML100470589 South Texas Project, Unit 3 & 4, 

Response to Request for Additional 

Information (RAI) Letter Number 10 

Related to the Application to Amend 

the ABWR DCD Part 2, Tier 2, 

Section 1.0 Provided in Attachment 1 

to the Referenced Letter. 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

2/18/2010 ML100550025 South Texas Project, Unit 3 & 4, 

Transmittal of Response to Request 

for Additional Information (RAI) Letter 

Numbers 2 and 5 Related to the 

Application to Amend the ABWR 

DCD Part 2, Tier 2, Section 19 

Provided in Attachment 1 to the 

Referenced Letter. 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

2/18/2010 ML100550027 South Texas Project, Units 3 & 4, 

Response to Request for Additional 

Information Letter No. 5, 6, 7, & 8, 

Related to the Application to Amend 

the ABWR DCD Part 2, Tier 2, 

Sections 3.8, 5.2, 15.8 and 19. 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 
B-2

Document 

Date 

ADAMS 

Accession 

Number 

Subject Correspondence From To Docket 

Number 

2/25/2010 ML100600410 South Texas Project, Units 3 & 4, 

Response to Request for Additional 

Information RAI Letter Numbers 5, 8, 

and 9 Related to the Application to 

Amend the ABWR DCD Part 2, Tier 

2, Sections 3.8, 9.2, and 19. 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

4/8/2010 ML101040254 South Texas Project, Units 3 & 4, 

Response to Request for Additional 

Information Letter Number 11 

Related to the Application to Amend 

the ABWR DCD Part 2, Tier 2, 

Section 14.2. 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

4/8/2010 ML101040345 South Texas Project, Units 3 & 4, 

Response to Request for Additional 

Information Related to Application to 

Amend ABWR DCD Part 2, Tier 2, 

Section 1.0 provided. 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

4/19/2010 ML101120085 South Texas Project, Units 3 and 4, 

Response to Request for Additional 

Information Question 09.02.04-1. 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

4/26/2010 ML101190120 South Texas Project, Units 3 and 4, 

Response to Request for Additional 

Information Question 03.02.02-1 

Related to Application Amend ABWR 

DCD, Part 2, Tier 2, Section 3.2. 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

5/12/2010 ML101340548 South Texas Project, Units 3 and 4, 

Proposed ABWR DCD AIA 

Ammendment Revision 1 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

7/21/2010 ML101470298 NRC Inspection Report Nos. 

05200012/2010201 and 

05200013/2010201 

Letter with 

Enclosures 

NRC, Quality 

and Vendor 

Branch 2, Office 

of New Reactor 

South Texas 

Project Nuclear 

Operating 

Company 

05200012 

05200013 

5/27/2010 ML101530610 South Texas Project, Units 3 and 4, 

Response to Request for Additional 

Information Question 03.08.04-2 and 

03.08.04-4 Related to Application to 

Amend ABWR DCD, Part 2, Tier 2, 

Section 3.2. 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

6/17/2010 ML101720306 South Texas Project, Units 3 and 4, 

STP ABWR DCD Aircraft Impact 

Assessment (AIA) Ammendment 

Revision 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

6/21/2010 ML101750070 South Texas Project, Units 3 and 4, 

Response to Request for Additional 

Information Letter Number 13 

Related to Application to Ammend 

ABWR DCD, Part 2, Tier 2, Section 

19. 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

7/12/2010 ML102000496 South Texas Project, Units 3 and 4, 

STP ABWR DCD Aircraft Impact 

Assessment (AIA) Ammendment 

Revision-0 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 
B-3

Document 

Date 

ADAMS 

Accession 

Number 

Subject Correspondence From To Docket 

Number 

8/4/2010 ML102240435 South Texas Project, Units 3 and 4, 

STP ABWR DCD Aircraft Impact 

Assessment (AIA) Ammendment 

Revision-1 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

9/2/2010 ML103190120 South Texas Project, Units 3 and 4, 

STP ABWR DCD Aircraft Impact 

Assessment (AIA) Ammendment 

Revision-2 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

9/23/2010 ML102870017 South Texas Project, Units 3 and 4, 

STP ABWR DCD Aircraft Impact 

Assessment (AIA) Ammendment 

Revision-3 

Letter South Texas 

Project Nuclear 

Operating 

Company 

NRC/Document 

Control Desk 

NRC/NRO 

05200001 

C-1

APPENDIX C 

PRINCIPAL CONTRIBUTORS 

Paul Kallan Senior Project Manager 

Nanette Gilles Senior Policy Analyst (AIA Working Group) 

Mark Caruso Senior Reliability and Risk Engineer (AIA Working Group) 

Bhagwat Jain Senior Structural Engineer (AIA Working Group) 

Dennis Andrukat General Engineer (AIA Working Group) 

David Jeng Senior Structural Engineer (AIA Working Group) 

Raj Goel Reactor Systems Engineer 

George Thomas Senior Reactor Engineer 

James Gilmer Reactor Systems Engineer 

Erick Martinez General Engineer 

Dinesh Taneja Senior Electronics Engineer 

Frank Talbot Operations Engineer 

Garrett Newman Reactor Operations Engineer 

Richard McIntyre Senior Reactor Engineer 

Angelo Stubbs Senior Reactor Systems Engineer 

Edward Fuller Senior Reliability and Risk Engineer 

Richard McNally Mechanical Engineer 

Mark Lintz Operations Engineer 

Peyton Doub Environmental Scientist 

Frank Akstulewicz Management Supervision 

Mark Tonacci Management Supervision 

Kimberly Hawkins Management Supervision 

Brent Clayton Management Supervision 

Lynn Mrowca Management Supervision 

John Segala Management Supervision 

John McKirgan Management Supervision 

Ian Jung Management Supervision 

Joseph Donoghue Management Supervision 

Richard Rasmussen Management Supervision 

Brian Thomas Management Supervision 

Jason Dreisbach Management Supervision 

Michael Junge Management Supervision 

Jennifer Dixon-Herrity Management Supervision 

Jerry Wilson Peer Review 

Tom Tai Peer Review 

Bernadette Abeywickrama Licensing Assistant 

Stacy Joseph Project Manager 

Contractor

Energy Research, Inc. SER Support

D-1

APPENDIX D 

CHRONOLOGY OF REQUESTS FOR ADDITIONAL INFORMATION (RAI) 

Question RAI RAI RAI Response 

 Number: SRP Section Title: Issued: Accession No: Response: Accession No: 

 

 01-2 01 - Introduction and Interfaces 1/13/10 ML100130817 2/8/10 ML100470589

 

 03.02.02-1 03.02.02 - System Quality Group Classification 1/26/10 ML100320305 3/3/10 ML100640162

 

 03.08.04-1 03.08.04 - Other Seismic Category I Structures 1/4/10 ML100050150 2/18/10 ML100550027

 

 03.08.04-2 03.08.04 - Other Seismic Category I Structures 1/4/10 ML100050150 5/27/10 ML101530610

 

 03.08.04-3 03.08.04 - Other Seismic Category I Structures 1/4/10 ML100050150 2/18/10 ML100550027

 

 03.08.04-4 03.08.04 - Other Seismic Category I Structures 1/4/10 ML100050150 5/27/10 ML101530610

 

 03.08.04-5 03.08.04 - Other Seismic Category I Structures 1/4/10 ML100050150 2/18/10 ML100550027

 

 05.02.02-1 05.02.02 - Overpressure Protection 12/18/09 ML093560920 1/20/10 ML100250139

 

 05.02.02-2 05.02.02 - Overpressure Protection 12/18/09 ML093560920 1/20/10 ML100250139

 

 05.02.02-3 05.02.02 - Overpressure Protection 1/4/10 ML100050050 2/18/10 ML100550027

 

 06.02.04-1 06.02.04 - Containment Isolation System 12/16/09 ML093500126 1/13/10 ML100190088

 

 07.07-1 07.07 - Control Systems 12/18/09 ML093560881 1/20/10 ML100250139

 

 07.07-2 07.07 - Control Systems 12/18/09 ML093560881 1/20/10 ML100250139

 

 07.07-3 07.07 - Control Systems 12/18/09 ML093560881 1/20/10 ML100250139

 

 09.02.04-1 09.02.04 - Potable and Sanitary Water Systems 1/13/10 ML100140773 4/26/10 ML101120085

 

 09.02.04-2 09.02.04 - Potable and Sanitary Water Systems 1/13/10 ML100140773 2/25/10 ML100600410

 

 09.02.04-3 09.02.04 - Potable and Sanitary Water Systems 1/13/10 ML100140773 2/25/10 ML100600410

 

 09.02.04-4 09.02.04 - Potable and Sanitary Water Systems 1/13/10 ML100140773 2/25/10 ML100600410

 

 14.02-1 14.02 - Initial Plant Test Program - Design Certification 1/14/10 ML100190138 8/4/10 ML101040254

 and New License Applicants 

 

 15.08-1 15.08 - Anticipated Transients Without Scram 12/30/09 ML093641061 2/18/10 ML100550027

 

 19-1 19 - Probabilistic Risk Assessment and Severe 12/18/09 ML093520830 2/18/10 ML100550025

 Accident Evaluation 

 

 19-2 19 - Probabilistic Risk Assessment and Severe 12/18/09 ML093520830 2/18/10 ML100550025

 Accident Evaluation 

 

 19-3 19 - Probabilistic Risk Assessment and Severe 12/18/09 ML093520830 2/18/10 ML100550025

 Accident Evaluation 

 

 19-4 19 - Probabilistic Risk Assessment and Severe 12/18/09 ML093520830 2/18/10 ML100550025

 Accident Evaluation 
D-2

 

Question RAI RAI RAI Response 

 Number: SRP Section Title: Issued: Accession No: Response: Accession No: 

 

 19-5 19 - Probabilistic Risk Assessment and Severe 12/18/09 ML093520830 2/18/10 ML100550025

 Accident Evaluation 

 

 19-6 19 - Probabilistic Risk Assessment and Severe 12/18/09 ML093520830 2/18/10 ML100550025

 Accident Evaluation 

 

 19-7 19 - Probabilistic Risk Assessment and Severe 12/18/09 ML093520830 2/18/10 ML100550025

 Accident Evaluation 

 

 19-8 19 - Probabilistic Risk Assessment and Severe 12/22/09 ML093641050 2/18/10 ML100550027

 Accident Evaluation 

 

 19-9 19 - Probabilistic Risk Assessment and Severe 12/22/09 ML093641050 2/18/10 ML100550027

 Accident Evaluation 

 

 19-10 19 - Probabilistic Risk Assessment and Severe 12/22/09 ML093641050 2/18/10 ML100550027

 Accident Evaluation 

 

 19-11 19 - Probabilistic Risk Assessment and Severe 12/22/09 ML093641050 2/25/10 ML100600410

 Accident Evaluation 

 

 19-12 19 - Probabilistic Risk Assessment and Severe 12/22/09 ML093641050 2/18/10 ML100550027

 Accident Evaluation 

 

 19-13 19 - Probabilistic Risk Assessment and Severe 12/22/09 ML093641050 2/18/10 ML100550025

 Accident Evaluation 

 

 19-14 19 - Probabilistic Risk Assessment and Severe 5/20/10 ML101410197 6/21/10 ML101750070

 Accident Evaluation 

 

 19-15 19 - Probabilistic Risk Assessment and Severe 5/20/10 ML101410197 6/21/10 ML101750070

 Accident Evaluation 
E-1

APPENDIX E 

REPORT BY ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
E-2
E-3
E-4
E-5
E-6
E-7

F-1

APPENDIX F

STAFF RESPONSE TO ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 
F-2
F-3
F-4
F-5
F-6
BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse)

NRC FORM 335

(12-2010)

NRCMD 3.7

U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

 (Assigned by NRC, Add Vol., Supp., Rev.,

 and Addendum Numbers, if any.)

NUREG - 1948

3. DATE REPORT PUBLISHED

MONTH

June

YEAR

2011

4. FIN OR GRANT NUMBER

2. TITLE AND SUBTITLE

Final Safety Evaluation Report Related to the Aircraft Impact Amendment to the U.S. Advanced

Boiling Water Reactor (ABWR) Design Certification

5. AUTHOR(S) 6. TYPE OF REPORT

Final Safety Evaluation

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address; if contractor,

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (If NRC, type "Same as above"; if contractor, provide NRC Division, Office or Region, U.S. Nuclear Regulatory Commission, 

provide name and mailing address.)

and mailing address.)

10. SUPPLEMENTARY NOTES

Docket No. 52-0001

11. ABSTRACT (200 words or less)

This Safety Evaluation Report (SER) documents the technical review of the U.S. Advanced Boiling Water Reactor (ABWR)

aircraft impact assessment (AIA) application by the U.S. Nuclear Regulatory Commission (NRC) staff. The STP Nuclear

Operating Company (STPNOC) submitted the application on June 30, 2009 to amend the design certification rule for the ABWR

in accorance with the procedures of Subpart B, "Standard Design Certifications" of Title 10 of the Code of Federal Regulations

(10 CFR) Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants". On September 23, 2010, STPNOC

submitted the final amendment application (Agencywide Documents Access and Management System Accession No.

ML102870017). The purpose of the amendment is to address the requirements in 10 CFR 50.150, "Aircraft Impact Assessment".

 If the proposed amendment is approved, applicants for a combined license that reference the ABWR standard design may

address the requirements of 10 CFR 50.150 by referencing the amended ABWR standard design. On the basis of its evaluation

and independent analyses, the NRC staff concludes that STPNOC's application meets the requirements of Subpart B of 10 CFR

Part 52 and 10 CFR 50.150(b).

12. KEY WORDS/DESCRIPTORS (List words or phrases that will assist researchers in locating the report.)

Aircraft Impact Assessment (AIA)

South Texas Project Nuclear Operating Company

Advanced Boiling Water Reactor (ABWR)

10 CFR 50.150

14. SECURITY CLASSIFICATION

13. AVAILABILITY STATEMENT

unlimited

(This Page)

unclassified

(This Report)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (12-2010)

Division of New Reactor Licensing, Office of New Reactors, U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

Same as above


UNITED STATES 

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

--------------------

OFFICIAL BUSINESS
NUREG-1948 Final Safety Evaluation Report Related to the Aircraft Impact Amendment to 

the U.S. Advanced Boiling Water Reactor (ABWR) Design Certification