Ennui I. Safety Elvaluation

Report

Related to Certification of the

A1'1 000 Standard Design

Supplement 1

U.' . Nuclear Regulatory Commission

Office of Nuclear Reactor Regulation

December 2005
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Final Safety Evaluation Report Related to Certification of the

AP1000 Standard Plant Design

Docket No. 52-006

Supplement 1

Division of New Reactor Licensing

Office of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

ABSTRACT

This report supplements the final safety evaluation report (FSER) for the AP1 000 standard

plant design. The FSER was issued by the U.S. Nuclear Regulatory Commission (NRC) as

NUREG-1793 in September 2004 to document the NRC staff's technical review of the AP1 000

design. The application for the AP1 000 design was submitted on June 28, 2002, by

Westinghouse Electric Corporation (Westinghouse) in accordance with Subpart B, "Standard

Design Certifications," of Part 52 of Title 10 of the Code of Federal Regulations (10 CFR

Part 52), and Appendix 0, "Standardization of Design: Staff Review of Standard Designs."

This supplement documents the NRC staff's review of Westinghouse's changes to the AP1000

design documentation in the design control document (DCD) since the issuance of the FSER.

On the basis of the evaluation described in the AP1000 FSER (NUREG-1793) and this report,

the NRC staff concludes that the changes to the DCD (up to and including Revision 15 to the

AP1 000 DCD) are acceptable and that Westinghouse's application for design certification

meets the requirements of Subpart B to 10 CFR Part 52 that are applicable and technically

relevant to the AP1000 standard plant design.

iii

TABLE OF CONTENTS

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

1. INTRODUCTION AND GENERAL DISCUSSION .1-1

1.1 Introduction .1-1

1.5 Summary of Principal Review Matters .1-2

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

5.4.1.2 Coast-Down Capability. 5-1

14. VERIFICATION PROGRAMS ........................... ...... 14-1

14.3 Tier 1 Information .. 14-1

15. TRANSIENT AND ACCIDENT ANALYSES . . .15-1

15.2.3.1 Partial Loss of Forced Reactor Coolant Flow . .15-1

24. CONCLUSION....................................................... 24-1

Table

Table 1.5-1 Summary of Changes to the APi000 DCD .1-3

v

1. INTRODUCTION AND GENERAL DISCUSSION

1.1 introduction

This report supplements the final safety evaluation report (FSER) for the AP1 000 standard

plant design. The U.S. Nuclear Regulatory Commission (NRC) staff issued the FSER as

NUREG-1 793 in September 2004 to document the NRC staff's review of the AP1 000. This

supplement documents the NRC staff's review of the changes to the AP1000 design

documentation since the issuance of the F-SER. Each section of this supplement is numbered

and titled the same as the section of the F:SER that is being updated. The discussions are

supplementary to, but not in lieu of, the discussions in the FSER, unless otherwise noted.

Westinghouse Electric Company, LLC (Westinghouse or the applicant) submitted the AP1000

design documentation under Subpart B of "Standard Design Certifications," of Part 52 of

Title 10 of the Code of Federal Regulations (10 CFR Part 52). The AP1 000 design

documentation includes the AP1 000 design control document (DCD) and probabilistic risk

assessment. Changes to the AP1 000 DCD (Docket No. 52-006) were submitted after the

FSER was issued. The staff's review of these DCD changes is discussed in Section 1.5 of this

report.

This supplement is issued by the Division of New Reactor Licensing in the Office of Nuclear

Reactor Regulation, NRC. The NRC's project managers for this part of the AP1000 design

certification review are Lauren M. Quinones-Navarro and Jerry N. Wilson, PE. They may be

reached by calling 301-415-2007 or 301-415-3145, or by writing to them at the Office of Nuclear

Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. The

AP1 000 design documentation and all revisions are available for public inspection at the NRC's

Public Document Room and the NRC's Public Electronic Reading Room (ADAMS). 1 The

NRC's Public Electronic Reading Room is at http:I/www.nrc.gov/reading-rm/adamslweb-based.html.

Through this Web site, the public can gain access to ADAMS, which provides text and image

files of NRC's public documents. The AP1000 FSER (NUREG-1793) and this supplement are

also available for public inspection at the NIRC's Public Document Room and Electronic

Reading Room.

1ADAMS (Agencywide Documents Access and Management System) is the NRC's information system that

provides access to all image and text documents that the NRC has made public since November 1, 1999, as well as

bibliographic records (some with abstracts and full text) that the NRC made public before November 1999.

Documents available to the public may be accessed via the Internet at

http://www.nrc.aov/readina-rm/adams/web-based.htrnl. Documents may also be viewed by visiting the NRC's Public

Document Room at One White Flint North, 11555 Rockville Pike, Rockville, Maryland. Telephone assistance for

using web-based ADAMS is available at (800) 397-4209 between 8:30 a.m. and 4:15 p.m., eastern standard time,

Monday through Friday, except Federal holidays.

1-1
1.5 Summary of Principal Review Matters

By letter dated September 7, 2005 (DCP/NRC 1722), Westinghouse submitted proposed Tier 1

AP1000 DCD changes. The NRC staff reviewed the proposed changes and provided

comments. Westinghouse submitted revised changes to the AP1 000 DCD by letters dated

November 1, 2005 (DCP/NRC 1723) and November 3, 2005 (DCP/NRC 1724), addressing the

staff's comments. Westinghouse incorporated the changes in Revision 15 of the DCD and the

errata to Revision 15 of the DCD submitted on November 14, 2005 (DCP/NRC 1725) and

December 8, 2005 (DCP/NRC 1727), respectively. The changes to the AP1 000 DCD include

editorial and minor technical changes and clarifications to the inspection, tests, analyses, and

acceptance criteria (ITAAC) in the Tier 1 information. Westinghouse identified these changes

as a result of the Nuclear Energy Institute (NEI) and NuStart efforts to prepare for future

combined license (COL) applications. These changes are listed in Table 1.5-1.

The NRC staff reviewed the changes, which were made to resolve inconsistencies between

Tier 1 and Tier 2 information in the AP1000 DCD, to correct administrative errors, and to make

editorial clarifications. The staff determined that most of these changes do not affect the staff's

findings in the AP1000 FSER and are acceptable. For some of the changes, a supplemental

evaluation is provided. Supplemental evaluations are set forth in Sections 5.4.1.2, 14.3,

and 15.2.3.1 of this supplement.

1-2
Table 1.5-1 Summary of Changes to the AP1000 DCD

TIER 1 SECTION DESCRIPTION

Section 1.1 Add tihe definition of "Tag Number" to the list of

definitions.

Section 1.2 Clarify the definition of "A report exists and concludes

that...".

Section 1.4 Delete the unnecessary line under "Safe Shutdown" in the

"Safe-.Shutdown Earthquake" definition on page 1.4-5 in

the Tier 1 "List of Acronyms and Abbreviations."

Table 2.1.1-1 Provide a precise reference for an ITAAC item.

Subsection 2.1.2 Correct several items in Tier 1, Tables 2.1.2-1 through

Tables 2.1.2-1 through 2.1.2-4 2.1.2-3. In Tier 1, Table 2.1.2-4, make an editorial

and Figures 2.1.2-1 and 2.1.2-2 correction in item 3.a, provide a precise reference for

ITAAC item 7.c, modify ITAAC item 8.b to verify the pump

coastclown curve rather than the calculated pump inertia,

and correct the pressurizer heater backup rated capacity

in item 9.b.

Table 2.1.2-1, and Tier 2 This change deletes the reactor coolant system (RCS)

Table 3.2-3, Table 3.11-1, flow velocity probe and replaces it with a flow signal

Subsections 5.4.3.2.1, 5.4.3.2.3, derived from the hot-leg elbow. References in Tier 2 text

7.2.1.1.3, Table 7.2-2, and tables to the cold-leg flow signals are also changed to

Subsections 9A.3.1.1.7, hot-leg signals to be consistent with Tier I changes.

9A.3.1.1.8, Table 9A-2, Chapter 15 figures related to analysis of partial loss of

Subsections 14.2.9.1.1, forced reactor coolant flow and the locked rotor event are

14.2.10.1.17, 14.2.10.4.11, also revised. The RCS loop layout, the RCS piping and

Table 1 5.0-4a, instrument diagram (P&ID), and the protection system

Subsections 15.3.1.1, 15.3.2.1, functional diagram of the flow instruments are revised.

Table 15.3-1, Figures 15.3.1-2

through 15.3.1-6,

Figures 15.3.3-3 through

15.3.3-7 and Technical

Specification (Section 16.1)

Table 3.3.1, Subsection 3.4.1

and Bases Subsections 3.3.1

and 3.4.1, Tier 2 Figure 5.1-3,

Figure 5.1-5 (Sheet 1 of 3),.and

Figure 7.2-1 (Sheet 5 of 20)

1-3
Subsection 2.1.3; Correct various tag numbers in the tables and make minor

Tables 2.1.3-1, 2.1.3-2, 2.1.3-3, editorial corrections. Provide a precise reference for an

2.1.3-4; and Figure 2.1.3-3, and ITAAC item in Tier 1, Table 2.1.3-2.

Tier 2 Tables 1.3-1 and 5.3-5;

Subsections 5.3.1.2 and 5.3.4.1;

and Figures 5.3-1 and 5.3-6

Tables 2.2.1-1, 2.2.1-2, and Change the tag numbers and the closure time for the

2.2.1-3; and Figure 2.2.1-1 containment purge isolation valves. Also provide a

precise reference for ITAAC item 6.c in Tier 1,

Table 2.2.1-3.

Subsection 2.2.2; Provide clarifications and corrections to Tier 1,

Tables 2.2.2-1, 2.2.2-2, and Subsection 2.2.2.

2.2.2-3; and Figure 2.2.2-1

Tables 2.2.3-1 through 2.2.3-4 Provide precise references for ITAAC items in the tables.

and Table 2.2.3-6; and Correct Tier 1, Figure 2.2.3-1 for the passive core cooling

Figure 2.2.3-1 system.

Tables 2.2.4-1, 2.2.4-2, 2.2.4-4, Revise certain tables in Tier 1, Subsection 2.2.4 to

and 2.2.4-5; and Figure 2.2.4-1 indicate that all main steam line steam generator pressure

sensors are qualified for a harsh environment and to

make a few other editorial corrections. Provide precise

references for ITAAC items 7.c and 8.c in Tier 1,

Table 2.2.4-4.

Subsection 2.2.5, Table 2.2.5-5, Correct certain tag numbers for equipment in the main

and Figure 2.2.5-1 control room emergency habitability system and make a

few additional editorial changes. Provide a precise

reference for ITAAC item 6.b in Table 2.2.5-5.

Table 2.3.1-2 Provide precise references for ITAAC Item 2 in Tier 1,

Table 2.3.1-2.

1-4
Subsection 2.3.2, and

Tables 2.3.2-1, 2.3.2-2, and

2.3.2-4

Correct a typographical error in the numbering of item 2 in

Tier 1, Subsection 2.3.2, "Design Description." In

Table 2.3.2-1, valve CVS-PL-V092 is erroneously shown

as being in a harsh environment. This valve is outside

containment and is not in a harsh environment. Revise

this valve entry in the table accordingly. In Table 2.3.2-2,

delete the pipeline L047 entry for the chemical and

volurme control system (CVS) letdown containment

penetration line. This was erroneously included in place

of L051. Add an entry for the piping to the pressurizer

auxiliary spray connection. This was erroneously omitted

from Table 2.3.2-2.

Expand Table 2.3.2-2 to include seismic analysis of all

CVS piping inside containment greater than 1-inch

nominal diameter and normally exposed to RCS pressure.

Add Note 1 to the table to indicate the extent to which

seismic analysis of lines with normally closed valves is

required. Also, in Table 2.3.2-2, correct several minor

typographical errors. In Table 2.3.2-4, provide precise

references for ITAAC items. Correct a typographical error

in itern 12.b.

Table 2.3.3-2 Correct the value for the volume of the ancillary diesel

generator fuel tank.

Table 2.3.4-2 and 2.3.4-4; and Provide precise references for ITAAC item 3 in Tier 1,

Tier 2 Figure 9.5.1-1 Table 2.3.4-2. Revise the list of fire protection system

piping that must remain functional following a

safe-shutdown earthquake. Revise Tier 2, Figure 9.5.1-1

to show line numbers.

Subsection 2.3.6; Update and/or correct normal residual heat removal

Tables 2.3.6-1, 2.3.6-2, and system (RNS) information in Tier 1, Subsection 2.3.6.

2.3.6-4; and Figure 2.3.6-1 and

Tier 2 Table 5.4-14

Subsection 2.3.7, and Correct and clarify Tier 1 information for the spent fuel

Tables 2.3.7-1, 2.3.7-2, 2.3.7-3, pool cooling system. Also provide precise references for

and 2.3.7-4 ITAAC items.

Figure 2.3.8-1 Correct an editorial error in Tier 1, Figure 2.3.8-1.

Tables 2.3.10-2 and 2.3.10-4 Provide precise references for ITAAC item 6.a) in Tier 1,

Table 2.3.10-4. Revise Table 2.3.10-4 item 7.a) and

Table 2.3.10-2 to correct omissions.

1-5
Subsection 2.3.13 and Make an editorial correction in Tier 1, Table 2.3.13-3 to

Table 2.3.13-3 indicate that there is only one check valve in the primary

sampling system. Provide precise references for ITAAC

items 6.c) and 7 in Tier 1, Table 2.3.13-3.

Table 2.3.14-2 Provide precise references for ITAAC item 2 in Tier 1,

Table 2.3.14-2.

Table 2.3.15-2 Provide precise references for ITAAC item 2 in Tier 1,

Table 2.3.15-2.

Table 2.5.2-8 Provide a precise reference for an ITAAC item.

Subsection 2.5.3 Revise the descriptions of the plant control system to

clarify that the AP1 000 plant control system uses software

signal selection rather than the hardware signal selection

used by the AP600.

Subsection 2.5.4 and Delete the following function of the data display and

Table 2.5.4-2 processing system (DDS):

"The DDS provides non-safety-related displays of

parameters originating in other systems."

Table 2.5.5-2 Provide a precise reference for an ITAAC item.

Table 2.6.1-4 Provide precise references for ITAAC items.

Table 2.6.3-3 Provide a precise reference for an ITAAC item.

Table 2.6.5-1 Provide a precise reference for an ITAAC item.

Table 2.7.1-4 and Table 2.7.1-5 Provide a precise reference for an ITAAC item in Tier 1,

Table 2.7.1-4. In addition, correct certain "Component

Name" entries and a "Tag No." entry in Tier 1,

Table 2.7.1-5.

Table 2.7.2-2 Provide precise references for ITAAC item 2 in Tier 1,

Table 2.7.2-2.

Figure 2.7.2-1 Correct certain tag numbers in Tier 1 Figure 2.7.2-1,

sheets 1 and 2 of 2.

Figure 2.7.4-1 Correct certain tag numbers in Tier 1 Figure 2.7.4-1,

sheets 1 and 2 of 2.

Table 2.7.5-3 Revise the "Component Location" for certain components

in Tier 1, Table 2.7.5-3.

Table 2.7.6-2 Provide precise references for ITAAC item 2 in Tier 1,

Table 2.3.1-2. Correct the units of a parameter from "cfm"

to "scfm."

1-6
Table 2.7.7-1 Revise the "Equipment Name" entries, a 'Tag No." entry,

and the "Display' entries in Tier 1, Table 2.7.7-1.

Table 3.2-1 Change the acceptance criteria to identify individual

criteria that could be completed in phases to facilitate

completion and review of the parts of the human factors

engineering program.

Subsection 3.3 and Tables 3.3-1 Revise Tier 1 Table 3.3-1 to correct and update the

and 3.3-6 dimensions of the concrete walls. Dimensions of the

concrete walls for the radwaste building were relocated to

Table 3.3-6.

Tables 3.5-1, 3.5-7, and 3.6-1 For Tier 1, Table 3.5-1, correct certain "Tag No." entries

and "Safety-Related Display" entries. For Tier 1,

Table 3.5-7, correct one "Component Location" entry.

Delete references to the specific technology used for

containment atmosphere radioactivity from the ITAAC for

reactor coolant pressure boundary leak detection. Also

provide precise references for the ITAAC electrical

separation item.

Table 5.0-1 Correct the description of the 'Tornado Missile Spectra" in

Table 5.0-1 of Tier 1.

1-7

5. REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS

5.4.1.2 Coast-Down Capability

In the last paragraph of Section 5.4.1.2 of the final safety evaluation report (FSER), the

U.S. Nuclear Regulatory Commission (NRC) staff stated that the acceptance criterion of the

reactor coolant pump (RCP) rotating inertia for flow coast-down capability specified in item 8.b

of Tier 1, Table 2.1.2-4, is no less than 695.3 kg-m2 (16,500 lb-ft). Westinghouse proposed to

change the acceptance criterion to a specific pump coast-down curve. This is done by (1)

adding Figure 2.1.2-2, "Flow Transient for Four Cold Legs in Operation, Four Pump

Coastdown," to Tier 1, Section 2.1.2; (2) changing the item 8.b entry of the "Inspections, Tests,

Analyses" column to read: "A test will be performed to determine the pump flow coast-down

curve"; and (3) changing the item 8.b entry of the "Acceptance Criteria" column to read: 'The

pump flow coast-down will provide RCS flows greater than or equal to the flow shown in

Figure 2.1.2-2, 'Flow Transient for Four Cold Legs in Operation, Four Pump Coastdown."'

Westinghouse proposed this change because the specified pump rotating inertia does not

guarantee that flows produced during pump coastdown will satisfy the analyses in the absence

of information on pump resistance. The proposed Figure 2.1.2-2 is a four-pump coast-down

curve identical to DCD Tier 2, Figure 15.3.2-1, in the safety analysis of the design-basis

transient of a complete loss of forced reactor coolant flow described in DCD Tier 2,

Subsection 15.3.2. The use of Tier 1, Figure 2.1.2-2, as the inspection, tests, analyses and

acceptance criteria (ITAAC) for the RCP flow coast-down capability ensures consistency with

the safety analysis and, therefore, the NRC staff finds such use acceptable.

Because of Westinghouse's change to the Tier 1 ITAAC acceptance criteria of the RCP pump

coastdown capability, the NRC is revising the last paragraph of Section 5.4.1.2 on page 5-57 of

NUREG-1793 to read as follows:

In DCD Tier 1, Table 2.1.2-4, "Inspections, Tests, Analyses, and Acceptance

Criteria," Item 8b specifies a test will be performed to determine the pump

coastdown curve that demonstrates the pump flow coastdown will provide RCS

flows greater than or equal to the flow shown in Tier 1 Figure 2.1.2-2, "Flow

Transient for Four Cold Legs in Operation, Four Pump Coastdown." Since the

four-pump coast-down curve in Tier 1, Figure 2.1.2-2 is identical to the four

pump coastdown curve in DCD Tier 2 Figure 15.3.2-1, which is used in the

safety analysis of the design basis transient of a complete loss of forced reactor

coolant flow described in DCD Tier 2, Subsection 15.3.2, and based on the

above evaluation, the staff finds the RCP coastdown capability curve acceptable

as an acceptance criterion for RCP capability.

5-1

14. VERIFICATION PROGRAMS

14.3 Tier 1 Information

In Section 14.3 of the final safety evaluation report, the NRC staff provided its evaluation of the

AP1000 inspections, tests, analyses, and acceptance criteria (ITAAC) in Tier 1 of the AP1000

design control document. In its submittals, dated November 1 and 3, 2005, Westinghouse

proposed many corrections and clarifications to the AP1 000 ITAAC, including some revisions to

the Tier 1 Introduction. The staff reviewed the proposed changes, performed an audit of the

AP1000 structural dimension changes as described below, and compared the revised structural

dimensions with those in the Tier 2 information. With the exception of the changes to wall

dimensions described below, the corrections and clarifications were not significant, and did not

affect the staff conclusions set forth in NUREG-1793.

In one of the Tier 1 changes, Westinghouse proposed a number of revisions to wall dimensions

(wall heights, wall lengths, and wall thicknesses) in Table 3.3-1, "Definition of Wall Thicknesses

for Nuclear Island Buildings and Annex Buildings," of the Tier 1 information. According to

Westinghouse, when Table 3.3-1 was initially developed from Tier 2 drawings, some wall

dimensions were not directly transferred. The purpose of this revision was to make the wall

dimensions in Table 3.3-1 consistent with those in the Tier 2 drawings.

The staff's review of Westinghouse's submittal raised the concern that if the revised Tier 1 wall

dimensions are not consistent with those of Tier 2 information, which were used for developing

the seismic model of the nuclear island buildings, the seismic responses calculated for the

design may be significantly affected. During conference calls, Westinghouse stated that the

seismic model was developed based on Tier 2 drawings and the revised Table 3.3-1 is now

consistent with those Tier 2 drawings. To verify Westinghouse's explanation, the staff

conducted an audit at Westinghouse's Rockville office on October 28, 2005, and confirmed that

the wall dimensions of the revised Table 3.3-1 are consistent with those in the Tier 2 design

drawings. On this basis, the staff concludes that the revised wall dimensions will not have any

effect on the seismic responses (design loads), and therefore, the changes to Tier 1

Table 3.3-1 are acceptable.

Based on the NRC staff's review of the AP1 000 Tier 1 information in accordance with the

guidance in SRP Section 14.3 (as described in Section 14.3 of the FSER) and its recent review

of Westinghouse's proposed corrections and clarifications to its Tier 1 information, described

above, the staff concludes that the top-level design features and performance characteristics of

the AP1 000 design are appropriately described in Tier 1 and the Tier 1 information is

acceptable. Further, the Tier 1 design descriptions can be adequately verified by ITAAC.

Therefore, the staff concludes that the ITAAC are necessary and sufficient to provide

reasonable assurance that, if the inspections, tests, and analyses are performed and the

acceptance criteria met, a facility referencing the certified design can be constructed and

operated in conformity with the design certification and the applicable regulations.

14-1

15. TRANSIENT AND ACCIDENT ANALYSES

15.2.3.1 Partial Loss of Forced Reactor Coolant Flow (DCD Tier 2. Section 15.3.1)

The first paragraph of Section 15.2.3.1 of the FSER states:

A mechanical or electrical failure in an RCP (reactor coolant pump], or a fault in

the power supply to the pumps supplied by an RCP bus, may cause partial loss

of RCS [reactor coolant system] flow, a moderate-frequency event. The low

primary coolant flow reactor trip signal in any reactor coolant loop provides

protection against this event.

Westinghouse proposed to change the RCS flow sensors from cold-leg velocity probes to hotleg elbow taps. The change document provided by Westinghouse states that experience with

cold-leg velocity probes has shown that there may be noise problems and signal-to-noise-ratio

problems with this type of flow sensor.

Since the AP1000 design has two hot loops and four cold loops, the NRC staff is replacing the

first paragraph of Section 15.2.3.1 on page 15-24 of NUREG-1793 with the following to remove

any ambiguity in the term "loop:"

A mechanical or electrical failure in an RCP, or a fault in the power supply to the

pumps supplied by an RCP bus, may cause partial loss of RCS flow, a

moderate-frequency event. The low' primary coolant flow reactor trip signal

provides protection against this event.

The deletion of the phrase "in any reactor coolant loop" is to reflect Westinghouse's design

change. The change in the location of the sensor will not impair its operation. Therefore, this

change does not affect our conclusions in Section 15.2.3.1 of the FSER.

15-1

24. CONCLUSION

The Nuclear Regulatory Commission (NRC) staff has reviewed Westinghouse's changes to the

AP1000 design documentation (see Section 1.5 of this report). On the basis of the evaluation

described in the AP1000 FSER (NUREG-1 793) and this report, the NRC staff concludes that

the AP 1000 design documentation (up to and including Revision 15 to the AP1 000 design

control document) is acceptable and that Westinghouse's application for design certification

meets the requirements of Subpart B, 'Standard Design Certifications," of Part 52 of Title 10 of

the Code of Federal Regulations (10 CFR Part 52) that are applicable and technically relevant

to the AP1000 standard plant design.

24-1

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(9-2004) (Assigned by NRC, Add Vol., Supp., Rev,

NRCMD 3.7 and Addendum Numbers, If any.)

BIBLIOGRAPHIC DATA SHEET

(See instnrctions on the reverse) NUREG-1793, Supplement 1

2. TITLE AND SUBTITLE 3. DATE REPORT PUBLISHED

Final Safety Evaluation Related to Certification of the AP1000 Standard Design, Supplement 1 MONTH YEAR

December 2005

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

Lauren Quinones-Navarro, Jerry N. Wilson, Wesley Held Safety Evaluation Report

7. PERIOD COVERED (Inclusi Dates)

8. PERFORMING ORGANIZATION -NAME AND ADDRESS (#NRC, pvide DMsin or Region. U.S NuciearR gulator Commissn anmading address; contractor.

provide name and maditxl address.)

Division of New Reactor Licensing

Oflice of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

9. SPONSORING ORGANIZATION -NAME AND ADDRESS (It NRC, type 'Sarne as above', icontractor, provide NRC Division. Office or Region, US. Nudear Regulatory Commission.

and maiing address.)

Same as above

10. SUPPLEMENTARY NOTES

Docket No. 52-006, Proiect No. 711

1 1 . ABSTRACT (200 word; orless)

This report supplements the final safety evaluation report (FS3ER) for the AP1 000 standard plant design. The FSER was issued

by the U.S. Nuclear Regulatory Commission (NRC) as NUREG-1 793 in September 2004 to document the NRC staff's technical

review of the AP1000 design. The application for the APIOGO design was submitted on June 28, 2002, by Westinghouse

Electric Corporation (Westinghouse) in accordance with Subpart B, Standard Design Certifications," of Part 52 of Title 10 of the

Code of Federal Regulations (10 CFR Part52), and Appendix O, "Standardization of Design: Staff Review of Standard

Designs." This supplement documents the NRC staff's review of Westinghouse's changes to the API 000 design documentation in the design control document (DCD) since the issuance of the FSER. On the basis of the evaluation

described in the AP1000 FSER (NUREG-1793) and this report, the NRC staff concludes that the changes to the DCD (up to

and including Revision 15 to the AP1 000 DCD) are acceptable and that Westinghouse's application for design certification

meets the requirements of Subpart B to 1 OCFR Part 52 that are applicable and technically relevant to the AP1 000 standard

plant design.

12. KEY WORDS/DESCRIPTORS (List words orphrases that will assist researchers A bcg the repoot) 13. AVAILABILITY STATEMENT

Advanced Reactor unlimited

Combined License 14. SECURITY CLASSIFICATION

Design Certification (This Page)

Final Design Approval unclassified

Inspections, Tests, Analyses, and Acceptance Criteria (shi Rwort)

Passive Plant Design unclassified

15. NUMBER OF PAGES

16. PRICE

N_ FOR PRIT_ __.924 _N .PAPER_ R. YCE _._ NRC FORM 335 (9-2004) PRINTED ON RECYCLED PAPER
Federal Recycling Program
I1
NUREG- 1793, Supp. 1 Final Safety Evaluation Report Related to Certification

of the AP1 000 Standard Design - Supplement 1 December 2005

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

OFFICIAL BUSINESS