RULEMAKING ISSUE

NOTATION VOTE

January 18, 2005 SECY-05-0017

FOR: The Commissioners

FROM: Luis A. Reyes

Executive Director for Operations

SUBJECT: PROPOSED RULE - AP1000 DESIGN CERTIFICATION

PURPOSE:

To obtain the Commission’s approval to publish in the Federal Register the attached proposed

rule that would amend 10 CFR Part 52 to certify the AP1000 standard plant design. 

BACKGROUND:

Westinghouse Electric Company (Westinghouse) submitted an application for certification of its

AP1000 standard plant design on March 28, 2002. The Nuclear Regulatory Commission (NRC)

staff issued a final design approval to Westinghouse on September 13, 2004, that signified

completion of the technical review phase and readiness for the rulemaking phase of the

AP1000 application. 

DISCUSSION:

The NRC staff completed its review of the AP1000 standard plant design and issued

NUREG-1793, “Final Safety Evaluation Report Related to Certification of the AP1000 Standard

Design,” dated November 2004. Certification of the AP1000 standard plant design will be

performed under Subpart B of 10 CFR Part 52.

This proposed design certification rule (DCR) is nearly identical to the previously approved DCR

for the AP600 design (Appendix C to 10 CFR Part 52) and therefore, the AP1000 DCR

emulates the existing DCR for the AP600 design. Also, Westinghouse requested that policy

resolutions for the AP600 be applied to the AP1000. Furthermore, many of the procedural

resolutions for the AP600 DCR (e.g., the two-tier structure, Tier 2*, and the scope of 

CONTACT: Lauren M. Quinones-Navarro, NRR/DRIP

 301-415-2007
The Commissioners -2-

issue resolution) were developed after extensive discussions with nuclear industry

representatives, including Westinghouse. However, there are some departures from the AP600

DCR, e.g., different design documentation, including Tier 2* information; design features; fewer

exemptions because of subsequent rule changes; and changes to the 50.59-like change

process to emulate the new 10 CFR 50.59 rule.

The attached Federal Register notice provides the public with an opportunity to comment on the

proposed DCR; the AP1000 Design Control Document (DCD), which is incorporated into the

DCR by reference; and the AP1000 environmental assessment. The time period for

commenting on the proposed AP1000 DCR is 75 days, consistent with the North American Free

Trade Agreement. Unlike the earlier DCRs, the proposed rule does not provide for an

opportunity to request a hearing, consistent with the newly-adopted changes to 10 CFR Parts 2

and 52 (69 FR 2182, January 14, 2004).

RECOMMENDATIONS:

That the Commission:

1. Approve the proposed amendment to 10 CFR Part 52 for publication in the Federal

Register.

2. Certify that this rule, if promulgated, will not have a negative economic impact on a

substantial number of small entities in order to satisfy requirements of the Regulatory

Flexibility Act, 5 U.S.C. 605(b).

3. Determine that the backfit rule, 10 CFR 50.109, does not apply to this proposed rule.

4. Note:

a. The proposed DCR will be published in the Federal Register for a 75-day comment

period;

b. An environmental assessment and finding of no significant impact has been prepared

(Attachment 2);

c. This proposed rule amends information collection requirements that are subject to the

Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). This rule will be submitted

to the Office of Management and Budget (OMB) for review and approval of the

paperwork requirements (Section V of Attachment 1);

d. The Chief Counsel for Advocacy of the Small Business Administration will be informed

of the certification regarding the economic impact on small entities and the reasons for it

as required by the Regulatory Flexibility Act (Section VII);

e. The appropriate congressional committees will be informed; and

f. The Office of Public Affairs will issue a press release. 
The Commissioners -3-

COORDINATION:

The Office of the General Counsel has reviewed this paper and has no legal objections. The

Chief Financial Officer has reviewed this paper for resource implications and has no objections. 

The Chief Information Officer has reviewed this paper for information technology and

information management implications and concurs in it. However, the proposed DCR requires

a change in the information collection requirements that will require a submission to the Office

of Management and Budget. A copy of the attached Federal Register notice was provided to

the Advisory Committee on Reactor Safeguards (ACRS) for its consideration. 

/RA Ellis W. Merschoff Acting For/

Luis A. Reyes

Executive Director

 for Operations

Attachments: 1. Federal Register notice

2. Environmental Assessment

3. Generic Cover Letter to State Liaison Officers
The Commissioners -3-

COORDINATION:

The Office of the General Counsel has reviewed this paper and has no legal objections. The

Chief Financial Officer has reviewed this paper for resource implications and has no objections. 

The Chief Information Officer has reviewed this paper for information technology and

information management implications and concurs in it. However, the proposed DCR requires

a change in the information collection requirements that will require a submission to the Office

of Management and Budget. A copy of the attached Federal Register notice was provided to

the Advisory Committee on Reactor Safeguards (ACRS) for its consideration. 

/RA Ellis W. Merschoff Acting For/

Luis A. Reyes

Executive Director

 for Operations

Attachments: 1. Federal Register notice

2. Environmental Assessment

3. Generic Cover Letter to State Liaison Officers

ADAMS Accession No. ML043230006-SECY Paper

*Via Memo

OFFICE RNRP RNRP Tech Editor RNRP RNRP:D

NAME LMQuinones JNWilson PKleene LADudes WDBeckner(LDudes for)

DATE 11/24/04 11/24/04 12/02/04 11/29/04 11/29/04

OFFICE DRIP:D STP OGC OE:D ADM:RDB

NAME DBMatthews(FGillespie for) PHLohaus(JM Piccone for) SATreby FJCongel(JGLuehman

for)

MTLesar*

DATE 12/21/04 12/06/04 12/20/04 12/03/04 12/08/04

OFFICE OCIO NRR:D EDO

NAME BJShelton* JEDyer LAReyes

DATE 12/20/04 01/04/05 01/16/05

OFFICIAL RECORD COPY
[7590-01-P]

NUCLEAR REGULATORY COMMISSION

10 CFR PART 52

RIN 3150-AH56

AP1000 Design Certification

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule. 

SUMMARY: The Nuclear Regulatory Commission (NRC or Commission) proposes to amend

its regulations to certify the AP1000 standard plant design. This action is necessary so that

applicants or licensees intending to construct and operate an AP1000 design may do so by

referencing the AP1000 design certification rule (DCR). This proposed DCR is nearly identical

to the AP600 DCR in the current regulations. The applicant for certification of the AP1000

design is Westinghouse Electric Company LLC (Westinghouse). The public is invited to submit

comments on this proposed DCR and the AP1000 design control document (DCD) that would

be incorporated by reference into the DCR. The NRC also invites the public to submit

comments on the environmental assessment for the AP1000 design.

DATE: Submit comments by [insert date 75 days after publication in the Federal Register.]

Comments received after this date will be considered if it is practical to do so, but the

Commission is able to ensure consideration only for comments received on or before this date.

ADDRESSES: You may submit comments by any one of the following methods. Please

include the following number (RIN 3150-AH56) in the subject line of your comments. 

Comments on rulemakings submitted in writing or in electronic form will be made available for

public inspection. Because your comments will not be edited to remove any identifying or

contact information, the NRC cautions you against including personal information such as social

security numbers and birth dates in your submission.
2

Mail comments to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, ATTN: Rulemakings and Adjudications Staff.

E-mail comments to: SECY@nrc.gov. If you do not receive a reply e-mail confirming

that we have received your comments, contact us directly at (301) 415-1966. You may also

submit comments via the NRC’s rulemaking web site at http://ruleforum.llnl.gov. Address

questions about our rulemaking website to Carol Gallagher (301) 415-5905; email cag@nrc.gov

Comments can also be submitted via the Federal eRulemaking Portal

http://www.regulations.gov. 

Hand deliver comments to: 11555 Rockville Pike, Rockville, Maryland 20852, between

the hours of 7:30 am and 4:15 pm Federal workdays (Telephone (301) 415-1966). 

Fax comments to: Secretary, U.S. Nuclear Regulatory Commission at (301) 415-1101.

Publicly available documents related to this rulemaking may be viewed electronically on

the public computers located at the NRC’s Public Document Room (PDR), O1 F21, One White

Flint North, 11555 Rockville Pike, Rockville, Maryland. The PDR reproduction contractor will

copy documents for a fee. Selected documents, including comments, can be viewed and

downloaded electronically via the NRC rulemaking web site at http://ruleforum.llnl.gov.

Publicly available documents created or received at the NRC after November 1, 1999,

are available electronically at the NRC's Electronic Reading Room at

http://www.nrc.gov/NRC/ADAMS/index.html. From this site, the public can gain entry into the

NRC's Agencywide Document Access and Management System (ADAMS), which provides text 
3

and image files of NRC's public documents. If you do not have access to ADAMS or if there

are problems in accessing the documents located in ADAMS, contact the NRC PDR Reference

staff at 1-800-397-4209, 301-415-4737, or by email to pdr@nrc.gov.

FOR FURTHER INFORMATION CONTACT: Lauren Quinones-Navarro or Jerry N. Wilson,

Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001; telephone (301) 415-2007 or (301) 415-3145; e-mail: lnq@nrc.gov or

jnw@nrc.gov. 

SUPPLEMENTARY INFORMATION:

TABLE OF CONTENTS

I. Background

II. Technical Evaluation of the AP1000 Design

III. Section-by-Section Discussion

A. Introduction (Section I)

B. Definitions (Section II)

C. Scope and Contents (Section III)

D. Additional Requirements and Restrictions (Section IV)

E. Applicable Regulations (Section V)

F. Issue Resolution (Section VI)

G. Duration of this Appendix (Section VII)

H. Processes for Changes and Departures (Section VIII)

I. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) (Section IX)

J. Records and Reporting (Section X)

IV. Availability of Documents
4

V. Plain Language

VI. Voluntary Consensus Standards

VII. Finding of No Significant Environmental Impact: Availability

VIII. Paperwork Reduction Act Statement

IX. Regulatory Analysis

X. Regulatory Flexibility Certification

XI. Backfit Analysis

XII. List of Subjects in 10 CFR Part 52

I. BACKGROUND

The NRC added 10 CFR Part 52 to its regulations to provide for the issuance of early

site permits (ESPs), standard design certifications, and combined licenses (COLs) for nuclear

power plants. Subpart B of 10 CFR Part 52 established the process for obtaining design

certifications. On March 28, 2002 (67 FR 20845), Westinghouse tendered its application for

certification of the AP1000 standard plant design with the NRC. Westinghouse submitted this

application in accordance with Subpart B and Appendix O of 10 CFR Part 52. The NRC

formally accepted the application as a docketed application for design certification (Docket

No. 52-006) on June 25, 2002 (67 FR 43690). The pre-application information submitted

before the NRC formally accepted the application can be found under Project No. 711. 

II. TECHNICAL EVALUATION OF THE AP1000 DESIGN

As stated above, the procedure for certifying a standard design is performed under

10 CFR Part 52, Subpart B, and is carried out in two stages (technical and administrative). The

technical review stage is initiated by an application filed in accordance with the requirements of

10 CFR 52.45, “Filing of Applications.” This stage continues with reviews by the NRC staff and
5

the Advisory Committee on Reactor Safeguards and ends with the issuance of a final safety

evaluation report (FSER) that discusses the staff’s conclusions related to the acceptability of

the AP1000 design. The NRC staff issued the AP1000 FSER in September 2004 

(NUREG-1793). The FSER provides the bases for issuance of a final design approval under

Appendix O to Part 52, which is a prerequisite to a design certification. The final design

approval for the AP1000 design was issued on September 13, 2004, and published in the

Federal Register on September 17, 2004 (69 FR 56101). 

The administrative review stage begins with the publication of a Federal Register notice

that initiates rulemaking, in accordance with 10 CFR 52.51, “Administrative Review of

Applications,” and includes a proposed design certification rule. The rulemaking culminates

with the denial of the application or the issuance of a design certification rule. 

III. SECTION-BY-SECTION DISCUSSION

The following discussion sets forth the purpose and key aspects of each section and

paragraph of the proposed AP1000 DCR. All section and paragraph references are to the

provisions in the proposed Appendix D to 10 CFR Part 52. The proposed DCR for the AP1000

standard plant design is nearly identical to the AP600 DCR, which the NRC previously codified

in 10 CFR Part 52, Appendix C (Design Certification Rule for the AP600 Design, 64 FR 72015,

December 23, 1999). Many of the procedural issues and their resolutions for the AP600 DCR

(e.g., the two-tier structure, Tier 2*, the scope of issue resolution) were developed after

extensive discussions with public stakeholders, including Westinghouse. Also, Westinghouse

requested that policy resolutions for the AP600 design review be applied to the AP1000. 

Accordingly, the NRC has modeled the AP1000 DCR on the existing DCRs, with certain

departures. These departures are necessary to account for differences in the AP1000 design 
6

documentation, design features, and environmental assessment (including severe accident

mitigation design alternatives). 

A. Introduction

The purpose of Section I of proposed Appendix D to 10 CFR Part 52 (this appendix)

would be to identify the standard plant design that is approved by this DCR and the applicant

for certification of the standard design. Identification of the design certification applicant is

necessary to implement this appendix, for two reasons. First, the implementation of 10 CFR

52.63 (c) depends on whether an applicant for a COL contracts with the design certification

applicant to provide the generic design control document (DCD) and supporting design

information. If the COL applicant does not use the design certification applicant to provide this

information, then the COL applicant must meet the requirements in 10 CFR 52.63 (c). Also,

X.A.1 of this appendix would impose a requirement on the design certification applicant to

maintain the generic DCD throughout the time period in which this appendix may be referenced.

B. Definitions

During development of the first two design certification rules, the Commission decided

that there would be both generic (master) DCDs maintained by the NRC and the design

certification applicant, as well as individual plant-specific DCDs, maintained by each applicant

and licensee who reference the appendix. This distinction is necessary in order to specify the

plant-specific requirements applicable to applicants and licensees referencing the appendix. 

The generic DCDs would reflect generic changes to the version of the DCD approved in this

design certification rulemaking. The generic changes would occur as the result of generic

rulemaking by the Commission, in accordance with the change criteria in Section VIII of this 
7

appendix. In addition, the Commission understood that each applicant and licensee referencing

this appendix would be required to submit and maintain a plant-specific DCD. 

This plant-specific DCD would contain (not just incorporate by reference) the information

in the generic DCD. The plant-specific DCD would be updated as necessary to reflect the

generic changes to the DCD that the Commission may adopt through rulemaking, any plantspecific departures from the generic DCD that the Commission imposed on the licensee by

order, and any plant-specific departures that the licensee chooses to make in accordance with

the relevant processes in Section VIII of this appendix. Thus, the plant-specific DCD would

function like an updated Final Safety Analysis Report (FSAR) because it would provide the most

complete and accurate information on a plant's licensing basis for that part of the plant within

the scope of this appendix. Therefore, this appendix would define both a generic DCD and a

plant-specific DCD. 

Also, the Commission decided to treat the technical specifications (TS) in Section 16.1

of the generic DCD as a special category of information and to designate them as generic TS in

order to facilitate the special treatment of this information under this appendix. A COL applicant

must submit plant-specific TS that consist of the generic TS, which may be modified under

paragraph VIII.C of this appendix, and the remaining plant-specific information needed to

complete the TS. The FSAR that is required by § 52.79(b) will consist of the plant-specific

DCD, the site-specific portion of the FSAR, and the plant-specific TS.

The terms Tier 1, Tier 2, Tier 2*, and COL action items (license information) are defined

in this appendix because these concepts were not envisioned when 10 CFR Part 52 was

developed. The design certification applicants and the NRC used these terms in implementing

the two-tiered rule structure that was proposed by representatives of the nuclear industry after

issuance of 10 CFR Part 52. Therefore, appropriate definitions for these additional terms are

included in this appendix. The nuclear industry representatives requested a two-tiered structure
8

for the design certification rules to achieve issue preclusion for a greater amount of information

than was originally planned for the design certification rules, while retaining flexibility for design

implementation. The Commission approved the use of a two-tiered rule structure in its staff

requirements memorandum (SRM), dated February 15, 1991, on SECY-90-377, “Requirements

for Design Certification Under 10 CFR Part 52,” dated November 8, 1990. This document and

others are available in the Regulatory History of Design Certification (see Section IV, Availability

of Documents). 

The Tier 1 portion of the design-related information contained in the DCD would be

certified by this appendix and, therefore, be subject to the special backfit provisions in

paragraph VIII.A of this appendix. An applicant who references this appendix would be

required to incorporate by reference and comply with Tier 1, under paragraphs III.B and IV.A.1

of this appendix. This information consists of an introduction to Tier 1, the system based and

non-system based design descriptions and corresponding inspections, tests, analyses, and

acceptance criteria (ITAAC), significant interface requirements, and significant site parameters

for the design. The design descriptions, interface requirements, and site parameters in Tier 1

were derived from Tier 2, but may be more general than the Tier 2 information. The NRC

staff's evaluation of the Tier 1 information is provided in Section 14.3 of the FSER. Changes to

or departures from the Tier 1 information must comply with Section VIII.A of this appendix. 

The Tier 1 design descriptions serve as commitments for the lifetime of a facility

referencing the design certification. The ITAAC verifies that the as-built facility conforms with

the approved design and applicable regulations. Under 10 CFR 52.103(g), the Commission

must find that the acceptance criteria in the ITAAC are met before authorizing operation. After

the Commission has made the finding required by 10 CFR 52.103(g), the ITAAC do not

constitute regulatory requirements for licensees or for renewal of the COL. However,

subsequent modifications to the facility must comply with the design descriptions in the plantspecific DCD unless changes are under the change process in Section VIII of this appendix. 
9

The Tier 1 interface requirements are the most significant of the interface requirements for

systems that are wholly or partially outside the scope of the standard design. Tier 1 interface

requirements were submitted in response to 10 CFR 52.47(a)(1)(vii) and must be met by the

site-specific design features of a facility that references this appendix. The Tier 1 site

parameters are the most significant site parameters, which were submitted in response to

10 CFR 52.47(a)(1)(iii). An application that references this appendix must demonstrate that the

site parameters (both Tier 1 and Tier 2) are met at the proposed site (refer to paragraph III.D of

this statement of consideration [SOC]). 

Tier 2 is the portion of the design-related information contained in the DCD that would

be approved by this appendix but not certified. Tier 2 information would be subject to the

backfit provisions in paragraph VIII.B of this appendix. Tier 2 includes the information required

by 10 CFR 52.47 (with the exception of generic TS, conceptual design information, and the

evaluation of severe accident mitigation design alternatives) and the supporting information on

inspections, tests, and analyses that will be performed to demonstrate that the acceptance

criteria in the ITAAC have been met. As with Tier 1, paragraphs III.B and IV.A.1 of this

appendix would require an applicant who references this appendix to incorporate Tier 2 by

reference and to comply with Tier 2, except for the COL action items, including the investment

protection short-term availability controls in Section 16.3 of the generic DCD. The definition of

Tier 2 makes clear that Tier 2 information has been determined by the Commission, by virtue of

its inclusion in this appendix and its designation as Tier 2 information, to be an approved

sufficient method for meeting Tier 1 requirements. However, there may be other acceptable

ways of complying with Tier 1. The appropriate criteria for departing from Tier 2 information

would be specified in paragraph VIII.B of this appendix. Departures from Tier 2 would not

negate the requirement in paragraph III.B to reference Tier 2. 
10

A definition of ?combined license action items” (COL information), which is part of the

Tier 2 information, would be added to clarify that COL applicants who reference this appendix

are required to address COL action items in their license application. However, the COL action

items are not the only acceptable set of information. An applicant may depart from or omit COL

action items, provided that the departure or omission is identified and justified in the FSAR. 

After issuance of a construction permit or COL, these items would not be requirements for the

licensee unless they are restated in the FSAR. For additional discussion, see Section D.

The investment protection short-term availability controls, which are set forth in

Section 16.3 of the generic DCD, would be added to the information that is part of Tier 2. 

These requirements were added to Tier 2 to make it clear that the availability controls are not

operational requirements for the purposes of paragraph VIII.C of this appendix. Rather, the

availability controls are associated with specific design features. The availability controls may

be changed if the associated design feature is changed under paragraph VIII.B of this

appendix. For additional discussion, see Section C.

Certain Tier 2 information has been designated in the generic DCD with brackets and

italicized text as "Tier 2*" information and, as discussed in greater detail in the section-bysection explanation for Section H, a plant-specific departure from Tier 2* information would

require prior NRC approval. However, the Tier 2* designation expires for some of this

information when the facility first achieves full power after the finding required by 10 CFR

52.103(g). The process for changing Tier 2* information and the time at which its status as

Tier 2* expires is set forth in paragraph VIII.B.6 of this appendix. Some Tier 2* requirements

concerning special preoperational tests are designated to be performed only for the first plant or

first three plants referencing the AP1000 DCR. The Tier 2* designation for these selected tests

would expire after the first plant or first three plants complete the specified tests. However, a

COL action item requires that subsequent plants shall also perform the tests or justify that the

results of the first-plant-only or first-three-plants-only tests are applicable to the subsequent
11

plant. 

In an earlier rulemaking (64 FR 53582; October 4, 1999), the Commission revised 

10 CFR 50.59 to incorporate new thresholds for permitting changes to a plant as described in

the FSAR without NRC approval. For consistency and clarity, the Commission proposes to use

these new thresholds in the proposed AP1000 DCR. Inasmuch as § 50.59 is the primary

change mechanism for operating nuclear plants, the Commission believes that future plants

referencing the AP1000 DCR should utilize thresholds as close to § 50.59 as is practicable and

appropriate. Because of some differences in how the change control requirements are

structured in the DCRs, certain definitions contained in § 50.59 are not applicable to 10 CFR

Part 52 and are not being included in this proposed rule. One definition that the Commission is

including in the definition from the new § 50.59 for a “departure from a method of evaluation,”

(paragraph II.G), which is appropriate to include in this rulemaking so that the eight criterion in

paragraph VIII.B.5.b of the proposed rule will be implemented as intended.

C. Scope and Contents

The purpose of Section III of this DCR would be to describe and define the scope and

contents of this design certification and to set forth how documentation discrepancies or

inconsistencies are to be resolved. Paragraph A is the required statement of the Office of the

Federal Register (OFR) for approval of the incorporation by reference of Tier 1, Tier 2, and the

generic TS into this appendix. Paragraph B requires COL applicants and licensees to comply

with the requirements of this appendix. The legal effect of incorporation by reference is that the

incorporated material has the same legal status as if it were published in the Code of Federal

Regulations. This material, like any other properly-issued regulation, has the force and effect of

law. Tier 1 and Tier 2 information, as well as the generic TS, have been combined into a single

document called the generic DCD, in order to effectively control this information and facilitate its

incorporation by reference into the rule. The generic DCD was prepared to meet the

requirements of the OFR for incorporation by reference (10 CFR Part 51). One of the
12

requirements of the OFR for incorporation by reference is that the design certification applicant

must make the generic DCD available upon request after the final rule becomes effective. 

Therefore, paragraph III.A of this appendix would identify a Westinghouse representative to be

contacted in order to obtain a copy of the generic DCD. 

Paragraphs A and B would also identify the investment protection short-term availability

controls in Section 16.3 of the generic DCD as part of the Tier 2 information. During its review

of the AP1000 design, the NRC determined that residual uncertainties associated with passive

safety system performance increased the importance of non-safety-related active systems in

providing defense-in-depth functions that back-up the passive systems. As a result,

Westinghouse developed administrative controls to provide a high level of confidence that

active systems having a significant safety role are available when challenged. Westinghouse

named these additional controls “investment protection short-term availability controls.” The

Commission included this characterization in Section III to ensure that these availability controls

are binding on applicants and licensees that reference this appendix and will be enforceable by

the NRC. The NRC’s evaluation of the availability controls is provided in Chapter 22 of the

FSER.

The generic DCD (master copy) for this design certification will be accessible

electronically in ADAMS and at the OFR. Copies of the generic DCD will also be available at

the NRC’s PDR. Questions concerning the accuracy of information in an application that

references this appendix will be resolved by checking the master copy of the generic DCD in

ADAMS. If a generic change (rulemaking) is made to the DCD by the change process provided

in Section VIII of this appendix, then at the completion of the rulemaking the NRC would

request approval of the Director, OFR, for the changed incorporation by reference and change

its copies of the generic DCD and notify the OFR and the design certification applicant to

change their copies. The Commission would require that the design certification applicant

maintain an up-to-date copy under paragraph X.A.1 of this appendix because it is likely that
13

most applicants intending to reference the standard design would obtain the generic DCD from

the design certification applicant. Plant-specific changes to and departures from the generic

DCD would be maintained by the applicant or licensee that references this appendix in a plantspecific DCD under paragraph X.A.2 of this appendix. 

In addition to requiring compliance with this appendix, paragraph B would clarify that the

conceptual design information and Westinghouse’s evaluation of severe accident mitigation

design alternatives are not considered to be part of this appendix. The conceptual design

information is for those portions of the plant that are outside the scope of the standard design

and are contained in Tier 2 information. As provided by 10 CFR 52.47(a)(1)(ix), these

conceptual designs are not part of this appendix and, therefore, are not applicable to an

application that references this appendix. Therefore, the applicant is not required to conform

with the conceptual design information that was provided by the design certification applicant. 

The conceptual design information, which consists of site-specific design features, was required

to facilitate the design certification review. Conceptual design information is neither Tier 1 nor

Tier 2. Section 1.8 of Tier 2 identifies the location of the conceptual design information. 

Westinghouse’s evaluation of various design alternatives to prevent and mitigate severe

accidents does not constitute design requirements. The Commission’s assessment of this

information is discussed in Section VII of this SOC on environmental impacts. 

Paragraphs C and D would set forth the manner in which potential conflicts would be

resolved. Paragraph C establishes the Tier 1 description in the DCD as controlling in the event

of an inconsistency between the Tier 1 and Tier 2 information in the DCD. Paragraph D would 
14

establish the generic DCD as the controlling document in the event of an inconsistency

between the DCD and the FSER for the certified standard design. 

Paragraph E would clarify that design activities that are wholly outside the scope of this

design certification may be performed using site-specific design parameters, provided the

design activities do not affect Tier 1 or Tier 2, or conflict with the interface requirements in the

DCD. This provision would apply to site-specific portions of the plant, such as the

administration building. Because this statement is not a definition, this provision has been

located in Section III of this appendix. 

D. Additional Requirements and Restrictions 

Section IV of this appendix would set forth additional requirements and restrictions

imposed upon an applicant who references this appendix. Paragraph IV.A would set forth the

information requirements for these applicants. This appendix would distinguish between

information and/or documents which must actually be included in the application or the DCD,

versus those which may be incorporated by reference (i.e., referenced in the application as if

the information or documents were included in the application). Any incorporation by reference

in the application should be clear and should specify the title, date, edition, or version of a

document, the page number(s), and table(s) containing the relevant information to be

incorporated.

Paragraph A.1 would require an applicant who references this proposed DCR to

incorporate by reference this DCR in its application. The legal effect of such an incorporation

by reference is that this appendix would be legally binding on the applicant or licensee. 

Paragraph A.2.a would require that a plant-specific DCD be included in the initial application. 

This would ensure that the applicant commits to complying with the DCD. This paragraph also
15

would require that the plant-specific DCD uses the same format as the generic DCD and

reflects the applicant’s proposed departures and exemptions from the generic DCD as of the

time of submission of the application. The Commission expects that the plant-specific DCD

would become the plant’s FSAR, by including information such as site-specific information for

the portions of the plant outside the scope of the referenced design, including related ITAAC,

and other matters required to be included in an FSAR by 10 CFR 50.34 and 52.79. Integration

of the plant-specific DCD and remaining site-specific information into the plant’s FSAR, would

result in an application that is easier to use and should minimize ?duplicate documentation” and

the attendant possibility for confusion. Paragraph A.2.a would also require that the initial

application include the reports on departures and exemptions as of the time of submission of

the application. 

Paragraph A.2.b would require that an application referencing this proposed DCR

include the reports required by paragraph X.B of this appendix for exemptions and departures

proposed by the applicant as of the date of submission of its application. Paragraph A.2.c

would require submission of plant-specific TS for the plant that consists of the generic TS from

Section 16.1 of the DCD, with any changes made under paragraph VIII.C of this appendix, and

the TS for the site-specific portions of the plant that are either partially or wholly outside the

scope of this design certification. The applicant must also provide the plant-specific information

designated in the generic TS, such as bracketed values. 

Paragraph A.2.d would require the applicant referencing this proposed DCR to provide

information demonstrating that the proposed site falls within the site parameters for this

appendix and that the plant-specific design complies with the interface requirements, as

required by 10 CFR 52.79(b). If the proposed site has a characteristic that exceeds one or

more of the site parameters in the DCD, then it would be unacceptable for this design unless 
16

the applicant seeks an exemption under Section VIII of this appendix and provides adequate

justification for locating the certified design on the proposed site. Paragraph A.2.e would

require submission of information addressing COL action items, identified in the generic DCD

as COL information in the application. The COL information identifies matters that need to be

addressed by an applicant who references this appendix, as required by Subpart C of 10 CFR

Part 52. An applicant may depart from or omit these items, provided that the departure or

omission is identified and justified in its application (FSAR). Paragraph A.2.f would require that

the application include the information specified by 10 CFR 52.47(a) that is not within the scope

of this rule, such as generic issues that must be addressed, in whole or in part, by an applicant

that references this rule. Paragraph A.3 would require the applicant to physically include, not

simply reference, the proprietary and safeguards information referenced in the DCD, or its

equivalent, to ensure that the applicant has actual notice of these requirements.

Paragraph IV.B would reserve the right to determine to the Commission in what manner

this DCR may be referenced by an applicant for a construction permit or operating license

under 10 CFR Part 50. This determination may occur in the context of a subsequent

rulemaking modifying 10 CFR Part 52 or this design certification rule, or on a case-by-case

basis in the context of a specific application for a 10 CFR Part 50 construction permit or

operating license. This provision is necessary because the previous DCRs were not

implemented in the manner that was originally envisioned at the time that 10 CFR Part 52 was

promulgated. The Commission’s concern is with the way ITAAC were developed and the lack

of experience with design certifications in license proceedings. Therefore, it is appropriate that

the Commission retain some discretion regarding the way this DCR could be referenced in a

10 CFR Part 50 licensing proceeding. 
17

E. Applicable Regulations 

The purpose of Section V of this appendix is to specify the regulations that would be

applicable and in effect if this proposed design certification is approved. These regulations

would consist of the technically relevant regulations identified in paragraph A, except for the

regulations in paragraph B that would not be applicable to this certified design. 

Paragraph A would identify the regulations in 10 CFR Parts 20, 50, 73, and 100 that are

applicable to the AP1000 design. The Commission’s determination of the applicable

regulations would be made as of the date specified in paragraph V.A of this appendix, which

would be the date that this appendix is approved by the Commission and signed by the

Secretary. 

In paragraph V.B of this appendix, the Commission would identify the regulations that do

not apply to the AP1000 design. The Commission has determined that the AP1000 design

should be exempt from portions of 10 CFR 50.34, 50.62, and Appendix A to Part 50, as

described in the FSER (NUREG-1793) and/or summarized below:

(1) Paragraph (f)(2)(iv) of 10 CFR 50.34 - Plant Safety Parameter Display Console.

Under 10 CFR 52.47(a)(ii), an applicant for design certification must demonstrate

compliance with any technically relevant Three Mile Island (TMI) requirements in 

10 CFR 50.34(f). The requirement in 10 CFR 50.34(f)(2)(iv) states that an application must

provide a plant safety parameter display console that will display a minimum set of parameters

defining the safety status of the plant, be capable of displaying a full range of important plant

parameters and data trends on demand, and be capable of indicating when process limits are

being approached or exceeded. Westinghouse addresses this requirement, in Section 18.8.2

of the DCD, with an integrated design rather than a stand-alone, add-on system, as is used at

most current operating plants. Specifically, Westinghouse integrated the safety parameter
18

display system (SPDS) requirements into the design requirements for the alarm and display

systems. The NRC staff has determined that the function of a separate SPDS may be

integrated into the overall control room design. Therefore, the Commission has determined that

the special circumstances for allowing an exemption as described in 10 CFR 50.12(a)(2)(ii)

exist because the requirement for an SPDS console need not be applied in this particular

circumstance to achieve the underlying purpose because Westinghouse has provided an

acceptable alternative that accomplishes the intent of the regulation. On this basis, the

Commission concludes that an exemption from the requirements of 10 CFR 50.34(f)(2)(iv) is

authorized by law, will not present an undue risk to public health and safety, and is consistent

with the common defense and security. 

(2) Paragraph (c)(1) of 10 CFR 50.62 - Auxiliary feedwater system.

The AP1000 design relies on the passive residual heat removal (PRHR) system in lieu

of an auxiliary or emergency feedwater system as its safety-related method of removing decay

heat. Westinghouse requested an exemption from a portion of 10 CFR 50.62(c)(1), which

requires auxiliary or emergency feedwater as an alternate system for decay heat removal

during an anticipated transient without scram (ATWS) event. The NRC staff concluded that

Westinghouse met the intent of the rule by relying on the PRHR system to remove the decay

heat and, thereby, met the underlying purpose of the rule. Therefore, the Commission has

determined that the special circumstances for allowing an exemption described in 10 CFR

50.12(a)(2)(ii) exist because the requirement for an auxiliary or emergency feedwater system is

not necessary to achieve the underlying purpose of 10 CFR 50.62(c)(1). This is because

Westinghouse has adopted acceptable alternatives that accomplish the intent of this regulation,

and the exemption is authorized by law, will not present an undue risk to public health and

safety, and is consistent with the common defense and security. 
19

(3) Appendix A to 10 CFR Part 50, GDC 17 - Offsite Power Sources.

Westinghouse requested a partial exemption from the requirement in General Design

Criteria (GDC) 17 for a second offsite power supply circuit. The AP1000 plant design supports

an exemption to this requirement by providing safety-related “passive” systems. These passive

safety-related systems only require electric power for valves and the related instrumentation. 

The onsite Class 1E batteries and associated dc and ac distribution systems can provide the

power for these valves and instrumentation. In addition, if no offsite power is available, it is

expected that the non-safety-related onsite diesel generators would be available for important

plant functions. However, this non-safety-related ac power is not relied on to maintain core

cooling or containment integrity. Therefore, the Commission has determined that the special

circumstances for allowing an exemption as described in 10 CFR 50.12(a)(2)(ii) exist because

the requirement need not be applied in this particular circumstance to achieve the underlying

purpose of having two offsite power sources. This is because the AP1000 design includes an

acceptable alternative approach to accomplish safety functions that do not rely on power from

the offsite system and, therefore, accomplishes the intent of the regulation. On this basis, the

Commission concludes that a partial exemption from the requirements of GDC 17 is authorized

by law, will not present an undue risk to public health and safety, and is consistent with the

common defense and security.

F. Issue Resolution 

The purpose of Section VI of this appendix would be to identify the scope of issues that

are resolved by the Commission in this rulemaking and; therefore, are ?matters resolved” within

the meaning and intent of 10 CFR 52.63(a)(4). The section is divided into five parts: (A) the

Commission’s safety findings in adopting this appendix, (B) the scope and nature of issues

which are resolved by this rulemaking, (C) issues which are not resolved by this rulemaking,
20

(D) the backfit restrictions applicable to the Commission with respect to this appendix, and

(E) the availability of secondary references.

Paragraph A would describe the nature of the Commission’s findings in general terms

and make the finding required by 10 CFR 52.54 for the Commission’s approval of this DCR. 

Furthermore, paragraph A would explicitly state the Commission’s determination that this

design provides adequate protection of the public health and safety.

Paragraph B would set forth the scope of issues that may not be challenged as a matter

of right in subsequent proceedings. The introductory phrase of paragraph B clarifies that issue

resolution as described in the remainder of the paragraph extends to the delineated NRC

proceedings referencing this appendix. The remainder of paragraph B describes the categories

of information for which there is issue resolution. Specifically, paragraph B.1 would provide that

all nuclear safety issues arising from the Atomic Energy Act of 1954, as amended, that are

associated with the information in the NRC staff’s FSER (NUREG-1793), the Tier 1 and Tier 2

information (including the availability controls in Section 16.3 of the generic DCD), and the

rulemaking record for this appendix are resolved within the meaning of § 52.63(a)(4). These

issues include the information referenced in the DCD that are requirements (i.e., ?secondary

references”), as well as all issues arising from proprietary and safeguards information which are

intended to be requirements. 

Paragraph B.2 would provide for issue preclusion of proprietary and safeguards

information. Paragraphs B.3, B.4, B.5, and B.6 would clarify that approved changes to and

departures from the DCD which are accomplished in compliance with the relevant procedures

and criteria in Section VIII of this appendix continue to be matters resolved in connection with

this rulemaking. Paragraphs B.4, B.5, and B.6, which would characterize the scope of issue

resolution in three situations, use the phrase ‘‘but only for that plant’’ (emphasis added). 
21

Paragraph B.4 would describe how issues associated with a design certification rule are

resolved when an exemption has been granted for a plant referencing the design certification

rule. Paragraph B.5 would describe how issues are resolved when a plant referencing the

design certification rule obtains a license amendment for a departure from Tier 2 information. 

Paragraph B.6 would describe how issues are resolved when the applicant or licensee

departs from the Tier 2 information on the basis of paragraph VIII.B.5, which would waive the

requirement to get NRC approval. In all three situations, after a matter (e.g., an exemption in

the case of paragraph B.4) is addressed for a specific plant referencing a design certification

rule, the adequacy of that matter for that plant would not ordinarily be subject to challenge in

any subsequent proceeding or action for that plant (such as an enforcement action) listed in the

introductory portion of paragraph IV.B. There would not, by contrast, be any issue resolution on

that subject matter for any other plant. 

Paragraph B.7 would provide that, for those plants located on sites whose site

parameters do not exceed those assumed in Westinghouse’s evaluation of severe accident

mitigation design alternatives (SAMDAs), all issues with respect to SAMDAs arising under the

National Environmental Policy Act of 1969 associated with the information in the environmental

assessment for this design and the information regarding SAMDAs in Appendix 1B of the

generic DCD are also resolved within the meaning and intent of § 52.63(a)(4). In the event an

exemption from a site parameter is granted, the exemption applicant has the initial burden of

demonstrating that the original SAMDA analysis still applies to the actual site parameters but; if

the exemption is approved, requests for litigation at the COL stage must meet the requirements

of § 2.309 and present sufficient information to create a genuine controversy in order to obtain

a hearing on the site parameter exemption.

Paragraph C would reserve the right of the Commission to impose operational

requirements on applicants that reference this appendix. This provision would reflect that
22

operational requirements, including generic TS in Section 16.1 of the DCD, were not completely

or comprehensively reviewed at the design certification stage. Therefore, the special backfit

provisions of § 52.63 do not apply to operational requirements. However, all design changes

would be controlled by the appropriate provision in Section VIII of this appendix. Although the

information in the DCD that is related to operational requirements was necessary to support the

NRC’s safety review of this design, the review of this information was not sufficient to conclude

that the operational requirements are fully resolved and ready to be assigned finality under

§ 52.63. As a result, if the NRC wanted to change a temperature limit and that operational

change required a consequential change to a design feature, then the temperature limit backfit

would be controlled by Section VIII (paragraph A or B) of this appendix. However, changes to

other operational issues, such as in-service testing and in-service inspection programs, postfuel load verification activities, and shutdown risk that do not require a design change would not

be restricted by § 52.63 (see VIII.C of this appendix). 

Paragraph C would allow the NRC to impose future operational requirements (distinct

from design matters) on applicants who reference this design certification. Also, license

conditions for portions of the plant within the scope of this design certification, e.g., start-up and

power ascension testing, are not restricted by § 52.63. The requirement to perform these

testing programs is contained in Tier 1 information. However, ITAAC cannot be specified for

these subjects because the matters to be addressed in these license conditions cannot be

verified prior to fuel load and operation, when the ITAAC are satisfied. Therefore, another

regulatory vehicle is necessary to ensure that licensees comply with the matters contained in

the license conditions. License conditions for these areas cannot be developed now because

this requires the type of detailed design information that will be developed during a combined

license review. In the absence of detailed design information to evaluate the need for and

develop specific post-fuel load verifications for these matters, the Commission is reserving the
23

right to impose license conditions by rule for post-fuel load verification activities for portions of

the plant within the scope of this design certification. 

Paragraph D would reiterate the restrictions (contained in Section VIII of this appendix)

placed upon the Commission when ordering generic or plant-specific modifications, changes or

additions to structures, systems, or components, design features, design criteria, and ITAAC

(VI.D.3 would address ITAAC) within the scope of the certified design. 

Paragraph E would provide the procedure for an interested member of the public to

obtain access to proprietary or safeguards information for the AP1000 design, in order to

request and participate in proceedings identified in paragraph VI.B of this appendix, viz.,

proceedings involving licenses and applications which reference this appendix. Paragraph E,

would specify that access must first be sought from the design certification applicant. If

Westinghouse refuses to provide the information, the person seeking access shall request

access from the Commission or the presiding officer, as applicable. Access to the proprietary

or safeguards information may be ordered by the Commission, but must be subject to an

appropriate non-disclosure agreement.

G. Duration of this appendix. 

The purpose of Section VII of this appendix would be in part, to specify the period during

which this design certification may be referenced by an applicant for a COL, under 10 CFR

52.55. This section would also state that the design certification would remain valid for an

applicant or licensee that references the design certification until the application is withdrawn or

the license expires. Therefore, if an application references this design certification during the

15-year period, then the design certification would be effective until the application is withdrawn

or the license issued on that application expires. Also, the design certification would be

effective for the referencing licensee if the license is renewed. The Commission intends for this
24

appendix to remain valid for the life of the plant that references the design certification to

achieve the benefits of standardization and licensing stability. This means that changes to or

plant-specific departures from information in the plant-specific DCD must be made under the

change processes in Section VIII of this appendix for the life of the plant. 

H. Processes for Changes and Departures. 

The purpose of Section VIII of this appendix would be to set forth the processes for

generic changes to or plant-specific departures (including exemptions) from the DCD. The

Commission adopted this restrictive change process in order to achieve a more stable licensing

process for applicants and licensees that reference this design certification rule. Section VIII is

divided into three paragraphs, which correspond to Tier 1, Tier 2, and operational requirements. 

The language of Section VIII distinguishes between generic changes to the DCD versus plantspecific departures from the DCD. Generic changes must be accomplished by rulemaking

because the intended subject of the change is the design certification rule itself, as is

contemplated by 10 CFR 52.63(a)(1). Consistent with 10 CFR 52.63(a)(2), any generic

rulemaking changes are applicable to all plants, absent circumstances which render the change

[?modification” in the language of § 52.63(a)(2)] ?technically irrelevant.” By contrast, plantspecific departures could be either a Commission-issued order to one or more applicants or

licensees; or an applicant or licensee-initiated departure applicable only to that applicant’s or

licensee’s plant(s), similar to a § 50.59 departure or an exemption. Because these

plant-specific departures will result in a DCD that is unique for that plant, Section X of this

appendix would require an applicant or licensee to maintain a plant-specific DCD. For purposes

of brevity, this discussion refers to both generic changes and plant-specific departures as

?change processes.”
25

Section VIII of this appendix and Section XI of this SOC refer to an ?exemption” from

one or more requirements of this appendix and the criteria for granting an exemption. The

Commission cautions that when the exemption involves an underlying substantive requirement

(applicable regulation), then the applicant or licensee requesting the exemption must also show

that an exemption from the underlying applicable requirement meets the criteria of 10 CFR

50.12.

Tier 1 information

The change processes for Tier 1 information would be covered in paragraph VIII.A. 

Generic changes to Tier 1 are accomplished by rulemaking that amends the generic DCD and

are governed by the standards in 10 CFR 52.63(a)(1). This provision provides that the

Commission may not modify, change, rescind, or impose new requirements by rulemaking

except when necessary either to bring the certification into compliance with the Commission’s

regulations applicable and in effect at the time of approval of the design certification or to

ensure adequate protection of the public health and safety or common defense and security. 

The rulemakings must provide for notice and opportunity for public comment on the proposed

change, as required by 10 CFR 52.63(a)(1). Departures from Tier 1 may occur in two ways:

(1) the Commission may order a licensee to depart from Tier 1, as provided in paragraph A.3;

or (2) an applicant or licensee may request an exemption from Tier 1, as provided in

paragraph A.4. If the Commission seeks to order a licensee to depart from Tier 1, paragraph

A.3 would require that the Commission find both that the departure is necessary for adequate

protection or for compliance, and that special circumstances are present. Paragraph A.4 would

provide that exemptions from Tier 1 requested by an applicant or licensee are governed by the

requirements of 10 CFR 52.63(b)(1) and 52.97(b), which provide an opportunity for a hearing. 
26

In addition, the Commission would not grant requests for exemptions that may result in a

significant decrease in the level of safety otherwise provided by the design. 

Tier 2 information

The change processes for the three different categories of Tier 2 information, namely,

Tier 2, Tier 2*, and Tier 2* with a time of expiration, would be set forth in paragraph VIII.B. The

change process for Tier 2 has the same elements as the Tier 1 change process, but some of

the standards for plant-specific orders and exemptions would be different. As stated in

Section III, of this preamble, it is the Commission’s intent that this appendix would emulate

Appendix C to 10 CFR Part 52. However, the Commission has revised the § 50.59-like change

process in paragraph VIII.B.5 of this appendix to be commensurate with the new 10 CFR 50.59

(64 FR 53613, October 4, 1994). 

The process for generic Tier 2 changes (including changes to Tier 2* and Tier 2* with a

time of expiration) tracks the process for generic Tier 1 changes. As set forth in paragraph B.1,

generic Tier 2 changes would be accomplished by rulemaking amending the generic DCD and

would be governed by the standards in 10 CFR 52.63(a)(1). This provision would provide that

the Commission may not modify, change, rescind, or impose new requirements by rulemaking

except when necessary, either to bring the certification into compliance with the Commission’s

regulations applicable and in effect at the time of approval of the design certification or to

ensure adequate protection of the public health and safety or common defense and security. If

a generic change is made to Tier 2* information, then the category and expiration, if necessary,

of the new information would also be determined in the rulemaking and the appropriate change

process for that new information would apply.
27

Departures from Tier 2 would occur in five ways: (1) the Commission may order a

plant-specific departure, as set forth in paragraph B.3; (2) an applicant or licensee may request

an exemption from a Tier 2 requirement as set forth in paragraph B.4; (3) a licensee may make

a departure without prior NRC approval under paragraph B.5 [the ?§ 50.59-like” process]; (4)

the licensee may request NRC approval for proposed departures which do not meet the

requirements in paragraph B.5 as provided in paragraph B.5.d; and (5) the licensee may

request NRC approval for a departure from Tier 2* information under paragraph B.6. 

Similar to Commission-ordered Tier 1 departures and generic Tier 2 changes,

Commission-ordered Tier 2 departures could not be imposed except when necessary either to

bring the certification into compliance with the Commission’s regulations applicable and in effect

at the time of approval of the design certification or to ensure adequate protection of the public

health and safety or common defense and security, as set forth in paragraph B.3. However, the

special circumstances for the Commission-ordered Tier 2 departures would not have to

outweigh any decrease in safety that may result from the reduction in standardization caused

by the plant-specific order, as required by 10 CFR 52.63(a)(3). The Commission determined

that it was not necessary to impose an additional limitation similar to that imposed on Tier 1

departures by 10 CFR 52.63(a)(3) and (b)(1). This type of additional limitation for

standardization would unnecessarily restrict the flexibility of applicants and licensees with

respect to Tier 2 information. 

An applicant or licensee would be permitted to request an exemption from Tier 2

information as set forth in proposed paragraph B.4. The applicant or licensee would have to

demonstrate that the exemption complies with one of the special circumstances in 10 CFR

50.12(a). In addition, the Commission would not grant requests for exemptions that may result

in a significant decrease in the level of safety otherwise provided by the design. However, the

special circumstances for the exemption do not have to outweigh any decrease in safety that
28

may result from the reduction in standardization caused by the exemption. If the exemption is

requested by an applicant for a license, the exemption would be subject to litigation in the same

manner as other issues in the license hearing, consistent with 10 CFR 52.63(b)(1). If the

exemption is requested by a licensee, then the exemption would be subject to litigation in the

same manner as a license amendment. 

For plant-specific Tier 2 information, the change process in the existing DCRs would be

commensurate with the change process in the former 10 CFR 50.59. The proposed rule would

revise paragraph VIII.B.5 to conform the terminology in the § 50.59-like change process to that

used in the revised § 50.59. This amendment would delete references to unreviewed safety

question and safety evaluation, and would conform to the evaluation criteria concerning when

prior NRC approval is needed. Also, a definition would be added (paragraph II.G) for

‘‘departure from a method of evaluation’’ to support the evaluation criterion in

paragraph VIII.B.5.b(8).

Paragraph B.5 would allow an applicant or licensee to depart from Tier 2 information,

without prior NRC approval, if the proposed departure does not involve a change to, or

departure from, Tier 1 or Tier 2* information, TS, or does not require a license amendment

under paragraphs B.5.b or B.5.c. The TS referred to in B.5.a of this paragraph are the TS in

Section 16.1 of the generic DCD, including bases, for departures made prior to issuance of the

COL. After issuance of the COL, the plant-specific TS would be controlling under

paragraph B.5. The bases for the plant-specific TS would be controlled by the bases control

procedures for the plant-specific TS (analogous to the bases control provision in the Improved

Standard Technical Specifications). The requirement for a license amendment in paragraph

B.5.b would be similar to the definition in the new 10 CFR 50.59 and apply to all information in

Tier 2 except for the information that resolves the severe accident issues. 
29

The Commission believes that the resolution of severe accident issues should be

preserved and maintain in the same fashion as all other safety issues that were resolved during

the design certification review (refer to SRM on SECY-90-377). However, because of the

increased uncertainty in severe accident issue resolutions, the Commission has proposed

separate criteria in paragraph B.5.c for determining if a departure from information that resolves

severe accident issues would require a license amendment. For purposes of applying the

special criteria in paragraph B.5.c, severe accident resolutions would be limited to design

features when the intended function of the design feature is relied upon to resolve postulated

accidents when the reactor core has melted and exited the reactor vessel, and the containment

is being challenged. The NRC is considering whether to create a definition using this

explanation for purposes of using paragraph B.5.c. These design features are identified in

Section 1.9.5 and Appendix 19B of the DCD, with other issues, and are described in other

sections of the DCD. Therefore, the location of design information in the DCD is not important

to the application of this special procedure for severe accident issues. However, the special

procedure in paragraph B.5.c would not apply to design features that resolve so-called “beyond

design basis accidents” or other low probability events. The important aspect of this special

procedure is that it would be limited to severe accident design features, as defined above. 

Some design features may have intended functions to meet ?design basis” requirements and to

resolve ?severe accidents.” If these design features are reviewed under paragraph VIII.B.5,

then the appropriate criteria from either paragraphs B.5.b or B.5.c would be selected depending

upon the function being changed. 

An applicant or licensee that plans to depart from Tier 2 information, under

paragraph VIII.B.5, would be required to prepare an evaluation which provides the bases for the

determination that the proposed change does not require a license amendment or involve a

change to Tier 1 or Tier 2* information, or a change to the TS, as explained above. In order to
30

achieve the Commission’s goals for design certification, the evaluation would need to consider

all of the matters that were resolved in the DCD, such as generic issue resolutions that are

relevant to the proposed departure. The benefits of the early resolution of safety issues would

be lost if departures from the DCD were made that violated these resolutions without

appropriate review. 

The evaluation of the relevant matters would need to consider the proposed departure

over the full range of power operation from startup to shutdown, as it relates to anticipated

operational occurrences, transients, design-basis accidents, and severe accidents. The

evaluation would also have to include a review of all relevant secondary references from the

DCD because Tier 2 information, which is intended to be treated as a requirement, would be

contained in the secondary references. The evaluation would consider Tables 14.3-1 through

14.3-8 and 19.59-18 of the generic DCD to ensure that the proposed change does not impact

Tier 1 information. These tables contain cross-references from the safety analyses and

probabilistic risk assessment in Tier 2 to the important parameters that were included in Tier 1. 

Although many issues and analyses could have been cross-referenced, the listings in these

tables were developed only for key analyses for the AP1000 design. 

A party to an adjudicatory proceeding (e.g., for issuance of a COL) who believes that an

applicant or licensee has not complied with paragraph VIII.B.5 when departing from Tier 2

information, would be permitted petition to admit such a contention into the proceeding under

paragraph B.5.f. This provision has been proposed because an incorrect departure from the

requirements of this appendix essentially would place the departure outside of the scope of the

Commission’s safety finding in the design certification rulemaking. Therefore, it follows that

properly founded contentions alleging such incorrectly implemented departures could not be

considered ?resolved” by this rulemaking. As set forth in paragraph B.5.f, the petition would

have to comply with the requirements of 10 CFR 2.309 and show that the departure does not
31

comply with paragraph B.5. Any other party would be allowed to file a response to the petition. 

If on the basis of the petition and any responses, the presiding officer in the proceeding

determines that the required showing has been made, the matter would be certified to the

Commission for its final determination. In the absence of a proceeding, petitions alleging

nonconformance with paragraph B.5 requirements applicable to Tier 2 departures would be 

treated as petitions for enforcement action under 10 CFR 2.206. 

Paragraph B.6 would provide a process for departing from Tier 2* information. The

creation of and restrictions on changing Tier 2* information resulted from the development of

the Tier 1 information for ABWR design certification (Appendix A to Part 52) and the 

ABB-CE System 80+ design certification (Appendix B to Part 52). During this development

process, these applicants requested that the amount of information in Tier 1 be minimized to

provide additional flexibility for an applicant or licensee who references these appendices. Also,

many codes, standards, and design processes, which would not be specified in Tier 1 that are

acceptable for meeting ITAAC, were specified in Tier 2. The result of these actions would be

that certain significant information only exists in Tier 2 and the Commission would not want this

significant information to be changed without prior NRC approval. This Tier 2* information

would be identified in the generic DCD with italicized text and brackets (See Table 1-1 of

AP1000 DCD Introduction). 

Although the Tier 2* designation was originally intended to last for the lifetime of the

facility, like Tier 1 information, the NRC determined that some of the Tier 2* information could

expire when the plant first achieves full (100 percent) power, after the finding required by

10 CFR 52.103(g), while other Tier 2* information must remain in effect throughout the life of

the facility. The factors determining whether Tier 2* information could expire after the first full

power was achieved were whether the Tier 1 information would govern these areas after first

full power and the NRC’s determination that prior approval was required before implementation
32

of the change due to the significance of the information. Therefore, certain Tier 2* information

listed in paragraph B.6.c would cease to retain its Tier 2* designation after full-power operation

is first achieved following the Commission finding under 10 CFR 52.103(g). Thereafter, that

information would be deemed to be Tier 2 information that would be subject to the departure

requirements in paragraph B.5. By contrast, the Tier 2* information identified in

paragraph B.6.b would retain its Tier 2* designation throughout the duration of the license,

including any period of license renewal. 

Certain preoperational tests in paragraph B.6.c would be designated to be performed

only for the first plant or first three plants that reference this appendix. Westinghouse’s basis

for performing these “first-plant-only” and “first-three-plants-only” preoperational tests is

provided in Section 14.2.5 of the DCD. The NRC found Westinghouse’s basis for performing

these tests and its justification for only performing the tests on the first plant or first three plants

acceptable. The NRC’s decision was based on the need to verify that plant-specific

manufacturing and/or construction variations do not adversely impact the predicted

performance of certain passive safety systems, while recognizing that these special tests would

result in significant thermal transients being applied to critical plant components. The NRC

believes that the range of manufacturing or construction variations that could adversely affect

the relevant passive safety systems would be adequately disclosed after performing the

designated tests on the first plant, or the first three plants, as applicable. The COL action item

in Section 14.4.6 of the DCD states that subsequent plants shall either perform these

preoperational tests or justify that the results of the first-plant-only or first-three-plant-only tests

are applicable to the subsequent plant. The Tier 2* designation for these tests would expire

after the first plant or first three plants complete these tests, as indicated in paragraph B.6.c. 

If Tier 2* information is changed in a generic rulemaking, the designation of the new

information (Tier 1, 2*, or 2) would also be determined in the rulemaking and the appropriate
33

process for future changes would apply. If a plant-specific departure is made from Tier 2*

information, then the new designation would apply only to that plant. If an applicant who

references this design certification makes a departure from Tier 2* information, the new

information would be subject to litigation in the same manner as other plant-specific issues in

the licensing hearing. If a licensee makes a departure from Tier 2* information, it would be

treated as a license amendment under 10 CFR 50.90 and the finality would be determined in

accordance with paragraph VI.B.5 of this appendix. Any requests for departures from Tier 2*

information that affects Tier 1 would also have to comply with the requirements in

paragraph VIII.A of this appendix. 

Operational Requirements

The change process for TS and other operational requirements in the DCD would be set

forth in paragraph VIII.C. This change process has elements similar to the Tier 1 and Tier 2

change process in paragraphs VIII.A and VIII.B, but with significantly different change

standards. Because of the different finality status for TS and other operational requirements

(refer to paragraph III.F of this SOC), the Commission decided to designate a special category

of information, consisting of the TS and other operational requirements, with its own change

process in proposed paragraph VIII.C. The key to using the change processes proposed in

Section VIII is to determine if the proposed change or departure would require a change to a

design feature described in the generic DCD. If a design change is required, then the

appropriate change process in paragraph VIII.A or VIII.B would apply. However, if a proposed

change to the TS or other operational requirements does not require a change to a design

feature in the generic DCD, then paragraph VIII.C would apply. The language in

paragraph VIII.C would also distinguish between generic (Section 16.1 of DCD) and plant-
34

specific TS to account for the different treatment and finality accorded TS before and after a

license is issued.

The process in proposed paragraph C.1 for making generic changes to the generic TS

in Section 16.1 of the DCD or other operational requirements in the generic DCD would be

accomplished by rulemaking and governed by the backfit standards in 10 CFR 50.109. The

determination of whether the generic TS and other operational requirements were completely

reviewed and approved in the design certification rulemaking would be based upon the extent to

which an NRC safety conclusion in the FSER is being modified or changed. If it cannot be

determined that the TS or operational requirement was comprehensively reviewed and finalized

in the design certification rulemaking, then there would be no backfit restriction under 10 CFR

50.109 because no prior position was taken on this safety matter. Some generic TS contain

bracketed values, which indicate that the NRC’s review is not complete. Generic changes

made under proposed paragraph VIII.C.1 would be applicable to all applicants or licensees

(refer to paragraph VIII.C.2), unless the change is irrelevant because of a plant-specific

departure. 

Plant-specific departures may occur by either a Commission order under proposed

paragraph VIII.C.3 or an applicant’s exemption request under paragraph VIII.C.4. The basis for

determining if the TS or operational requirement was completely reviewed and approved for

these processes would be the same as for proposed paragraph VIII.C.1 above. If the TS or

operational requirement is comprehensively reviewed and finalized in the design certification

rulemaking, then the Commission must demonstrate that special circumstances are present

before ordering a plant-specific departure. If not, there would be no restriction on plant-specific

changes to the TS or operational requirements, prior to the issuance of a license, provided a

design change is not required. Although the generic TS were reviewed by the NRC staff to

facilitate the design certification review, the Commission intends to consider the lessons
 1 For discussion of the verification of ITAAC, see SECY-00-0092, “Combined License Review

Process,” dated April 20, 2000. 

35

learned from subsequent operating experience during its licensing review of the plant-specific

TS. The process for petitioning to intervene on a TS or operational requirement would be

similar to other issues in a licensing hearing, except that the petitioner must also demonstrate

why special circumstances are present (paragraph VIII.C.5). 

Finally, the generic TS would have no further effect on the plant-specific TS after the

issuance of a license that references this appendix. The bases for the generic TS would be

controlled by the change process in paragraph VIII.C of this appendix. After a license is issued,

the bases would be controlled by the bases change provision set forth in the administrative

controls section of the plant-specific TS. 

I. Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) 

The purpose of Section IX of this appendix would be to set forth how the ITAAC in Tier 1

of this design certification rule would to be treated in a license proceeding. Paragraph A would 

restate the responsibilities of an applicant or licensee for performing and successfully

completing ITAAC, and notifying the NRC of such completion. Paragraph A.1 would clarify that

an applicant may proceed at its own risk with design and procurement activities subject to

ITAAC, and that a licensee may proceed at its own risk with design, procurement, construction,

and preoperational testing activities subject to an ITAAC, even though the NRC may not have

found that any particular ITAAC has been successfully completed. Paragraph A.2 would

require the licensee to notify the NRC that the required inspections, tests, and analyses in the

ITAAC have been completed and that the acceptance criteria have been met. 

Paragraphs B.1 and B.2 would reiterate the NRC’s responsibilities with respect to ITAAC

as set forth in 10 CFR 52.99 and 52.103(g)1

. Finally, paragraph B.3 would state that ITAAC do
36

not, by virtue of their inclusion in the DCD, constitute regulatory requirements after the licensee

has received authorization to load fuel or has been granted a renewal of its license. However,

subsequent modifications to the terms of the COL would have to comply with the design

descriptions in the DCD unless the applicable requirements in 10 CFR 52.97 and Section VIII of

this appendix have been met. As discussed in paragraph III.D of this SOC, the Commission

would defer a determination of the applicability of ITAAC and its effect in terms of issue

resolution in 10 CFR Part 50 licensing proceedings to such time that a Part 50 applicant

decides to reference this appendix. 

J. Records and Reporting 

The purpose of Section X of this appendix would be to set forth the requirements that

would apply to maintaining records of changes to and departures from the generic DCD, which

would be reflected in the plant-specific DCD. Section X also would set forth the requirements

for submitting reports (including updates to the plant-specific DCD) to the NRC. This section of

the appendix would be similar to the requirements for records and reports in 10 CFR Part 50,

except for minor differences in information collection and reporting requirements. 

Paragraph X.A.1 of this appendix would require that a generic DCD and the proprietary and

safeguards information referenced in the generic DCD be maintained by the applicant for this

rule. The generic DCD was developed, in part, to meet the requirements for incorporation by

reference, including availability requirements. Therefore, the proprietary and safeguards

information could not be included in the generic DCD because they are not publicly available. 

However, the proprietary and safeguards information was reviewed by the NRC and, as stated

in proposed paragraph VI.B.2 of this appendix, the Commission would consider the information

to be resolved within the meaning of 10 CFR 52.63(a)(4). Because this information is not in the

generic DCD, the proprietary and safeguards information, or its equivalent, would be required to
37

be provided by an applicant for a license. Therefore, to ensure that this information will be

available, a requirement for the design certification applicant to maintain the proprietary and

safeguards information was added to proposed paragraph X.A.1 of this appendix. The

acceptable version of the proprietary and safeguards information would be identified

(referenced) in the version of the DCD that would be incorporated into this rule. The generic

DCD and the acceptable version of the proprietary and safeguards information would be

maintained for the period of time that this appendix may be referenced. 

Paragraphs A.2 and A.3 would place recordkeeping requirements on the applicant or

licensee that references this design certification so that its plant-specific DCD accurately

reflects both generic changes to the generic DCD and plant-specific departures made under

proposed Section VIII of this appendix. The term ?plant-specific” would be added to

paragraph A.2 and other sections of this appendix to distinguish between the generic DCD that

would be incorporated by reference into this appendix, and the plant-specific DCD that the

applicant would be required to submit under proposed paragraph IV.A of this appendix. The

requirement to maintain the generic changes to the generic DCD would be explicitly stated to

ensure that these changes are not only reflected in the generic DCD, which would be

maintained by the applicant for design certification, but that the changes would also be reflected

in the plant-specific DCD. Therefore, records of generic changes to the DCD would be required

to be maintained by both entities to ensure that both entities have up-to-date DCDs. 

Paragraph X.A of this appendix would not place recordkeeping requirements on sitespecific information that is outside the scope of this rule. As discussed in paragraph III.D of this

SOC, the FSAR required by 10 CFR 52.79 would contain the plant-specific DCD and the

site-specific information for a facility that references this rule. The phrase ?site-specific portion

of the final safety analysis report” in paragraph X.B.3.c of this appendix would refer to the

information that is contained in the FSAR for a facility (required by 10 CFR 52.79) but is not part
38

of the plant-specific DCD (required by proposed paragraph IV.A of this appendix). Therefore,

this rule would not require that duplicate documentation be maintained by an applicant or

licensee that references this rule, because the plant-specific DCD would be part of the FSAR

for the facility.

Paragraph X.B.1 would require applicants or licensees that reference this rule to submit

reports, which describe departures from the DCD and include a summary of the written

evaluations. The requirement for the written evaluations would be set forth in paragraph X.A.1. 

The frequency of the report submittals would be set forth in paragraph X.B.3. The requirement

for submitting a summary of the evaluations would be similar to the requirement in 10 CFR

50.59(d)(2). 

 Paragraph X.B.2 would require applicants or licensees that reference this rule to submit

updates to the DCD, which include both generic changes and plant-specific departures. The

frequency for submitting updates would be set forth in paragraph X.B.3. The requirements in

paragraph X.B.3 for submitting the reports and updates would vary according to certain time

periods during a facility’s lifetime. If a potential applicant for a combined license who

references this rule decides to depart from the generic DCD prior to submission of the

application, then paragraph B.3.a would require that the updated DCD be submitted as part of

the initial application for a license. Under proposed paragraph B.3.b, the applicant may submit

any subsequent updates to its plant-specific DCD along with its amendments to the application

provided that the submittals are made at least once per year. Because amendments to an

application are typically made more frequently than once a year, this should not be an

excessive burden on the applicant. 

Paragraph B.3.b would also require that the reports required by paragraph X.B.1 be

submitted semi-annually. This increase in reporting frequency during the period of construction

and application review is consistent Commission guidance. Also, more frequent reporting of
39

design changes during the period of detailed design and construction is necessary to closely

monitor the status and progress of the facility. In order to make the finding under 10 CFR

52.103(g), the NRC must monitor the design changes made under proposed Section VIII of this

appendix. Frequent reporting of design changes would be particularly important in times when

the number of design changes could be significant, such as during the procurement of

components and equipment, detailed design of the plant before and during construction, and

during preoperational testing. After the facility begins operation, the frequency of reporting

would revert to the requirement in paragraph B.3.c, which is consistent with the requirements

for plants licensed under 10 CFR 50.57. 

IV. AVAILABILITY OF DOCUMENTS

The NRC is making the documents identified below available to interested persons

through one or more of the following:

Public Document Room (PDR). The NRC’s Public Document Room is located at

11555 Rockville Pike, Public File Area O-1 F21, Rockville, MD 20082. Copies of publicly

available documents related to this rulemaking can be viewed electronically on public

computers in the PDR. The PDR reproduction contractor will make copies of documents for a

fee.

Rulemaking Website (Web). The NRC’s interactive rulemaking Website is located at

http://ruleforum.llnl.gov. Selected documents may be viewed and downloaded electronically via

this Website.

Public Electronic Reading Room (ADAMS). The NRC’s public Electronic Reading Room

is located at http://www.nrc.gov/reading-rm/adams.html. Through this site, the public can gain

access to ADAMS, which provides text and image files of NRC’s public documents.
 2 The regulatory history of the NRC’s design certification reviews is a package of 100 documents that

is available in NRC’s (PERR) and in the PDR. This history spans a 15-year period during which the

NRC simultaneously developed the regulatory standards for reviewing these designs and the form and

content of the rules that certified the designs.

40

Note: Public access to documents, including access via ADAMS and the PDR, has been

temporarily suspended so that security reviews of publicly available documents may be

performed and potentially sensitive information removed. However, access to the

documents identified in this rule continues to be available through the rulemaking web

site at http://ruleforum.llnl.gov, which was not affected by the ADAMS shutdown. Please

check with the listed NRC contact concerning any issues related to document

availability.

Document PDR Web ADAMS

AP1000 Design Certification Proposed Rule SECY paper x x ML043230006

AP1000 Environmental Assessment x x ML043230023

AP1000 Design Control Document x ......

NUREG-1793, “AP1000 Final Safety Evaluation Report” x ......

SECY-99-268, “Final Rule- AP600 Design Certification” x ...... ML003708259

Regulatory History of Design Certification2

x ...... ML003761550

V. PLAIN LANGUAGE

The Presidential memorandum entitled ?Plain Language in Government Writing”

(63 FR 31883; June 10, 1998), directed that the Government’s writing be in plain language. 

The NRC requests comments on the proposed rule specifically with respect to the clarity and

effectiveness of the language used. Comments should be sent using one of the methods

detailed under the ADDRESSES heading of the preamble to this proposed rule.

VI. VOLUNTARY CONSENSUS STANDARDS
41

The National Technology and Transfer Act of 1995 (Act), Public Law 104-113, requires

that Federal agencies use technical standards that are developed or adopted by voluntary

consensus standards bodies unless using such a standard is inconsistent with applicable law or

is otherwise impractical. In this proposed rule, the NRC proposes to approve the AP1000

standard plant design for use in a combined license (COL) application under 10 CFR Part 52 or

possibly for a construction permit (CP) application under 10 CFR Part 50. Design certifications

are not generic rulemakings establishing a generally applicable standard with which all Parts 50

and 52 nuclear power plant licensees must comply. Design certifications are Commission

approvals of specific nuclear power plant designs by rulemaking. Furthermore, design

certification rulemakings are initiated by an applicant for rulemaking, rather than by the NRC. 

For these reasons, the NRC concludes that the act does not apply to this proposed rule. 

VII. FINDING OF NO SIGNIFICANT ENVIRONMENTAL IMPACT: AVAILABILITY

The Commission has determined under the National Environmental Policy Act of 1969,

as amended (NEPA), and the Commission’s regulations in 10 CFR Part 51, Subpart A, that this

proposed design certification rule, if adopted, would not be a major Federal action significantly

affecting the quality of the human environment and, therefore, an environmental impact

statement (EIS) is not required. The basis for this determination, as documented in the

environmental assessment, is that this amendment to 10 CFR Part 52 would not authorize the

siting, construction, or operation of a facility using the AP1000 design; it would only codify the

AP1000 design in a rule. The NRC will evaluate the environmental impacts and issue an EIS

as appropriate under NEPA as part of the application(s) for the construction and operation of a

facility. 
42

In addition, as part of the environmental assessment for the AP1000 design, the NRC

reviewed Westinghouse’s evaluation of various design alternatives to prevent and mitigate

severe accidents in Appendix 1B of the AP1000 DCD Tier 2. Based upon review of

Westinghouse’s evaluation, the Commission finds that: (1) Westinghouse identified a

reasonably complete set of potential design alternatives to prevent and mitigate severe

accidents for the AP1000 design; (2) none of the potential design alternatives are justified on

the basis of cost-benefit considerations; and (3) it is unlikely that other design changes would

be identified and justified in the future on the basis of cost-benefit considerations, because the

estimated core damage frequencies for the AP1000 are very low on an absolute scale. Finally,

the Commission concludes that the AP1000 design would not exclude severe accident

mitigation design alternatives for a future facility that would prove cost beneficial had they been

considered as part of the original design certification application. These issues are considered

resolved for the AP1000 design.

The environmental assessment (EA), upon which the Commission’s finding of no

significant impact is based, and the AP1000 DCD are available for examination and copying at

the NRC Public Document Room, One White Flint North, 11555 Rockville Pike, Rockville,

Maryland. The NRC has sent a copy of the EA and this proposed rule to every State Liaison

Officer and request their comments on the EA. Single copies of the EA are also available from

Lauren M. Quinones-Navarro, Mailstop O-4D9A, Office of Nuclear Reactor Regulation, U.S.

Nuclear Regulatory Commission, Washington, DC 20555. 

VIII. PAPERWORK REDUCTION ACT STATEMENT

This proposed rule contains amended information collection requirements that are

subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq). This rule has been 
43

submitted to the Office of Management and Budget for review and approval of the information

collection requirements.

Type of submission, new or revision: Revision

The title of the information collection: Appendix D to 10 CFR Part 52, AP1000 Design

 Certification, Proposed Rule (X.B.3.b)

The form number if applicable: Not applicable

How often the collection is required: Semi-annually

Who will be required or asked to report: Electric power companies who referenced a

design certification rule for a nuclear power plant. 

An estimate of the number of annual responses: 2 

The estimated number of annual respondents: 1 (applicant for a combined license) 

An estimate of the total number of hours needed annually to complete the requirement

or request: 16 hours (8 hours per response) 

Abstract: The NRC is proposing to amend its regulations to certify the AP1000 standard

plant design under Subpart B of 10 CFR Part 52. This action is necessary so that applicants or

licensees intending to construct and operate an AP1000 design may do so by referencing the

AP1000 design certification rule (DCR). This proposed DCR, as set out in Appendix D, is

nearly identical to the AP600 DCR in Appendix C of 10 CFR Part 52. The applicant for

certification of the AP1000 design is Westinghouse Electric Company LLC. 

The U.S. Nuclear Regulatory Commission is seeking public comment on the potential

impact of the information collection contained in this proposed rule and on the following issues: 

1. Is the proposed information collection necessary for the proper

performance of the functions of the NRC, including whether the

information will have practical utility?

2. Is the estimate of burden accurate?
44

3. Is there a way to enhance the quality, utility, and clarity of the information

to be collected?

4. How can the burden of the information collection be minimized, including

the use of automated collection techniques?

A copy of the OMB clearance package may be viewed free of charge at the NRC Public

Document Room, One White Flint North, 11555 Rockville Pike, Room O-1 F21, Rockville, MD

20852. The OMB clearance package and rule are available at the NRC worldwide Web site:

http://www.nrc.gov/public-involve/doc-comment/omb/index.html for 60 days after the signature

date of this notice and are also available at the rule forum site, http://ruleforum.llnl.gov.

Send comments on any aspect of these proposed information collections, including

suggestions for reducing the burden and on the above issues, by (INSERT DATE 30 DAYS

AFTER PUBLICATION IN THE FEDERAL REGISTER) to the Records and FOIA/Privacy

Services Branch (T-5 F52), U.S. Nuclear Regulatory Commission, Washington, DC 20555-

0001, or by Internet electronic mail to INFOCOLLECTS@NRC.GOV and to the Desk Officer,

John A. Asalone, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0151),

Office of Management and Budget, Washington, DC 20503. Comments received after this date

will be considered if it is practical to do so, but assurance of consideration cannot be given to

comments received after this date. You may also e-mail comments to

John_A._Asalone@omb.eop.gov or comment by telephone at (202) 395-4650.

Public Protection Notification

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

request for information or an information collection requirement unless the requesting

document displays a currently valid OMB control number.
45

IX. REGULATORY ANALYSIS

The NRC has not prepared a regulatory analysis for this proposed rule. The NRC

prepares regulatory analyses for rulemakings that establish generic regulatory requirements

applicable to all licensees. Design certifications are not generic rulemakings in the sense that

design certifications do not establish standards or requirements with which all licensees must

comply. Rather, design certifications are Commission approvals of specific nuclear power plant

designs by rulemaking, which then may be voluntarily referenced by applicants for COLs. 

Furthermore, design certification rulemakings are initiated by an applicant for a design

certification, rather than the NRC. Preparation of a regulatory analysis in this circumstance

would not be useful because the design to be certified is proposed by the applicant rather than

the NRC. For these reasons, the Commission concludes that preparation of a regulatory

analysis is neither required nor appropriate.

X. REGULATORY FLEXIBILITY CERTIFICATION

Under the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the Commission certifies

that this proposed rulemaking will not have a significant economic impact upon a substantial

number of small entities. This proposed rule provides for certification of a nuclear power plant

design. Neither the design certification applicant, nor prospective nuclear power plant licensees

who reference this design certification rule, fall within the scope of the definition of ?small

entities” set forth in the Regulatory Flexibility Act, or the Small Business Size Standards set out

in regulations issued by the Small Business Administration in 13 CFR Part 121. Thus, this rule

does not fall within the purview of the act.
46

XI. BACKFIT ANALYSIS

The Commission has determined that this proposed rule does not constitute a

backfitting as defined in the backfit rule, 10 CFR 50.109 because this design certification does

not impose new or changed requirements on existing 10 CFR Part 50 licensees, nor does it

impose new or change requirements on existing DCRs in Appendices A-C of Part 52. 

Therefore, a backfit analysis was not prepared for this rule. 

XII. LIST OF SUBJECTS IN 10 CFR PART 52

Part 52 - Administrative practice and procedure, Antitrust, Backfitting, Combined license,

Early site permit, Emergency planning, Fees, Incorporation by reference, Inspection, Limited

work authorization, Nuclear power plants and reactors, Probabilistic risk assessment,

Prototype, Reactor siting criteria, Redress of site, Reporting and record keeping requirements,

Standard design, Standard design certification.

For the reasons set out in the preamble and under the authority of the Atomic Energy

Act of 1954, as amended; the Energy Reorganization Act of 1974, as amended; and

5 U.S.C. 553; the NRC is proposing to adopt the following amendment to 10 CFR Part 52.

1. The authority citation for 10 CFR Part 52 continues to read as follows:

AUTHORITY: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 936, 948, 953, 954,

955, 956, as amended, sec. 234, 83 Stat. 444, as amended (42 U.S.C. 2133, 2201, 2232,

2233, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended

(42 U.S.C. 5841, 5842, 5846); sec. 1704, 112 Stat. 2750 (44 U.S.C. 3504 note). 

2. In § 52.8, paragraph (b) is revised to read as follows:

§ 52.8 Information collection requirements: OMB approval.
 3

AP1000 is a trademark of Westinghouse Electric Company LLC.

47

*****

(b) The approved information collection requirements contained in this part appear in

§§ 52.15, 52.17, 52.29, 52.35, 52.45, 52.47, 52.51, 52.57, 52.63, 52.75, 52.77, 52.78, 52.79,

52.89, 52.91, 52.99, and appendices A, B, C, and D to this point. 

3. A new Appendix D to 10 CFR Part 52 is added to read as follows:

Appendix D To Part 52 - Design Certification Rule for the AP1000 Design

I. INTRODUCTION

Appendix D constitutes the standard design certification for the AP10003

 design, in

accordance with 10 CFR Part 52, Subpart B. The applicant for certification of the AP1000

design is Westinghouse Electric Company LLC. 

II. DEFINITIONS

A. Generic design control document (generic DCD) means the document containing the

Tier 1 and Tier 2 information and generic TS that is incorporated by reference into this

appendix.

B. Generic technical specifications means the information required by 10 CFR 50.36

and 50.36a for the portion of the plant that is within the scope of this appendix.

C. Plant-specific DCD means the document maintained by an applicant or licensee who

references this appendix consisting of the information in the generic DCD as modified and

supplemented by the plant-specific departures and exemptions made under Section VIII of this
48

appendix.

D. Tier 1 means the portion of the design-related information contained in the generic

DCD that is approved and certified by this appendix (Tier 1 information). The design

descriptions, interface requirements, and site parameters are derived from Tier 2 information. 

Tier 1 information includes:

1. Definitions and general provisions; 

2. Design descriptions; 

3. Inspections, tests, analyses, and acceptance criteria (ITAAC); 

4. Significant site parameters; and 

5. Significant interface requirements. 

E. Tier 2 means the portion of the design-related information contained in the generic

DCD that is approved but not certified by this appendix (Tier 2 information). Compliance with

Tier 2 is required, but generic changes to and plant-specific departures from Tier 2 are

governed by Section VIII of this appendix. Compliance with Tier 2 provides a sufficient, but not

the only acceptable, method for complying with Tier 1. Compliance methods differing from

Tier 2 must satisfy the change process in Section VIII of this appendix. Regardless of these

differences, an applicant or licensee must meet the requirement in Paragraph III.B to reference

Tier 2 when referencing Tier 1. Tier 2 information includes:

1. Information required by 10 CFR 52.47, with the exception of generic TS and

conceptual design information; 

2. Information required for a final safety analysis report under 10 CFR 50.34;

3. Supporting information on the inspections, tests, and analyses that will be performed

to demonstrate that the acceptance criteria in the ITAAC have been met; and

4. COL action items (COL information), which identify certain matters that shall be

addressed in the site-specific portion of the FSAR by an applicant who references this
49

appendix. These items constitute information requirements but are not the only acceptable set

of information in the FSAR. An applicant may depart from or omit these items, provided that

the departure or omission is identified and justified in the FSAR. After issuance of a

construction permit or COL, these items are not requirements for the licensee unless such

items are restated in the FSAR. 

5. The investment protection short-term availability controls in Section 16.3 of the DCD.

F. Tier 2* means the portion of the Tier 2 information, designated as such in the generic

DCD, which is subject to the change process in paragraph VIII.B.6 of this appendix. This

designation expires for some Tier 2* information under paragraph VIII.B.6. 

G. Departure from a method of evaluation described in the plant-specific DCD used in

establishing the design bases or in the safety analyses means:

1. Changing any of the elements of the method described in the plant-specific DCD

unless the results of the analysis are conservative or essentially the same; or

2. Changing from a method described in the plant-specific DCD to another method

unless that method has been approved by the NRC for the intended application. 

H. All other terms in this appendix have the meaning set out in 10 CFR 50.2, 10 CFR

52.3, or Section 11 of the Atomic Energy Act of 1954, as amended, as applicable. 

III. SCOPE AND CONTENTS

A. Tier 1, Tier 2 (including the investment protection short-term availability controls in

Section 16.3), and the generic TS in the AP1000 DCD (Revision 14) are approved for

incorporation by reference by the Director of the Office of the Federal Register on [Insert date

of approval] under 5 U.S.C. 552(a) and 1 CFR Part 51. Copies of the generic DCD may be

obtained from Ronald P. Vijuk, Manager, Passive Plant Engineering, Westinghouse Electric

Company, P.O. Box 355, Pittsburgh, PA 15230-0355. A copy of the generic DCD is also
50

available for examination and copying at the NRC Public Document Room, One White Flint

North, 11555 Rockville Pike, Rockville, Maryland. Copies are available for examination at the

NRC Library, 11545 Rockville, Maryland, telephone (301) 415-5610, e-mail

LIBRARY@NRC.GOV or at the National Archives and Records Administration (NARA). For

information on the availability of this material at NARA, call (202) 741-6030 or go to

http://www.archives.gov/federal_register/code_of_federal_regulations/ibr_locations.html.

B. An applicant or licensee referencing this appendix, in accordance with Section IV of

this appendix, shall incorporate by reference and comply with the requirements of this

appendix, including Tier 1, Tier 2 (including the investment protection short-term availability

controls in Section 16.3 of the DCD), and the generic TS except as otherwise provided in this

appendix. Conceptual design information in the generic DCD and the evaluation of severe

accident mitigation design alternatives in Appendix 1B of the generic DCD are not part of this

appendix. 

C. If there is a conflict between Tier 1 and Tier 2 of the DCD, then Tier 1 controls. 

D. If there is a conflict between the generic DCD and either the application for design

certification of the AP1000 design or NUREG-1793, ?Final Safety Evaluation Report Related to

Certification of the AP1000 Standard Design,” (FSER), then the generic DCD controls. 

E. Design activities for structures, systems, and components that are wholly outside the

scope of this appendix may be performed using site-specific design parameters, provided the

design activities do not affect the DCD or conflict with the interface requirements. 

IV. ADDITIONAL REQUIREMENTS AND RESTRICTIONS

A. An applicant for a license that wishes to reference this appendix shall, in addition to

complying with the requirements of 10 CFR 52.77, 52.78, and 52.79, comply with the following

requirements: 
51

1. Incorporate by reference, as part of its application, this appendix.

2. Include, as part of its application:

a. A plant-specific DCD containing the same information and utilizing the same

organization and numbering as the AP1000 DCD, as modified and supplemented by the

applicant’s exemptions and departures;

b. The reports on departures from and updates to the plant-specific DCD required by

paragraph X.B of this appendix;

c. Plant-specific TS, consisting of the generic and site-specific TS that are required by

10 CFR 50.36 and 50.36a;

d. Information demonstrating compliance with the site parameters and interface

requirements;

e. Information that addresses the COL action items; and

f. Information required by 10 CFR 52.47(a) that is not within the scope of this appendix.

3. Physically include, in the plant-specific DCD, the proprietary and safeguards

information referenced in the AP1000 DCD. 

B. The Commission reserves the right to determine in what manner this appendix may

be referenced by an applicant for a construction permit or operating license under Part 50. 

V. APPLICABLE REGULATIONS

A. Except as indicated in paragraph B of this section, the regulations that apply to the

AP1000 design are in 10 CFR Parts 20, 50, 73, and 100, codified as of [insert date final rule

signed], that are applicable and technically relevant, as described in the FSER (NUREG-1793).

B. The AP1000 design is exempt from portions of the following regulations:

1. 10 CFR 50.34(f)(2)(iv) - Plant Safety Parameter Display Console;

2. 10 CFR 50.62(c)(1) - Auxiliary (or emergency) feedwater system; and
52

3. 10 CFR Part 50, Appendix A, GDC 17 - Offsite Power Sources.

VI. ISSUE RESOLUTION

A. The Commission has determined that the structures, systems, components, and

design features of the AP1000 design comply with the provisions of the Atomic Energy Act of

1954, as amended, and the applicable regulations identified in Section V of this appendix; and

therefore, provide adequate protection to the health and safety of the public. A conclusion that

a matter is resolved includes the finding that additional or alternative structures, systems,

components, design features, design criteria, testing, analyses, acceptance criteria, or

justifications are not necessary for the AP1000 design.

B. The Commission considers the following matters resolved within the meaning of

10 CFR 52.63(a)(4) in subsequent proceedings for issuance of a COL, amendment of a COL,

or renewal of a COL, proceedings held under to 10 CFR 52.103, and enforcement proceedings

involving plants referencing this appendix:

1. All nuclear safety issues, except for the generic TS and other operational

requirements, associated with the information in the FSER, Tier 1, Tier 2 (including referenced

information, which the context indicates is intended as requirements, and the investment

protection short-term availability controls in Section 16.3 of the DCD), and the rulemaking

record for certification of the AP1000 design;

2. All nuclear safety and safeguards issues associated with the information in

proprietary and safeguards documents, referenced and in context, are intended as

requirements in the generic DCD for the AP1000 design;

3. All generic changes to the DCD under and in compliance with the change processes

in Sections VIII.A.1 and VIII.B.1 of this appendix;

4. All exemptions from the DCD under and in compliance with the change processes in
53

Sections VIII.A.4 and VIII.B.4 of this appendix, but only for that plant;

5. All departures from the DCD that are approved by license amendment, but only for

that plant;

6. Except as provided in paragraph VIII.B.5.f of this appendix, all departures from Tier 2

under and in compliance with the change processes in paragraph VIII.B.5 of this appendix that

do not require prior NRC approval, but only for that plant;

7. All environmental issues concerning severe accident mitigation design alternatives

(SAMDAs) associated with the information in the NRC’s EA for the AP1000 design and

Appendix 1B of the generic DCD, for plants referencing this appendix whose site parameters

are within those specified in the SAMDA evaluation. 

C. The Commission does not consider operational requirements for an applicant or

licensee who references this appendix to be matters resolved within the meaning of 10 CFR

52.63(a)(4). The Commission reserves the right to require operational requirements for an

applicant or licensee who references this appendix by rule, regulation, order, or license

condition. 

D. Except under the change processes in Section VIII of this appendix, the Commission

may not require an applicant or licensee who references this appendix to:

1. Modify structures, systems, components, or design features as described in the

generic DCD;

2. Provide additional or alternative structures, systems, components, or design features

not discussed in the generic DCD; or

3. Provide additional or alternative design criteria, testing, analyses, acceptance criteria,

or justification for structures, systems, components, or design features discussed in the generic

DCD. 

E.1. Persons who wish to review proprietary and safeguards information or other
54

secondary references in the AP1000 DCD, in order to request or participate in the hearing

required by 10 CFR 52.85 or the hearing provided under 10 CFR 52.103, or to request or

participate in any other hearing relating to this appendix in which interested persons have

adjudicatory hearing rights, shall first request access to such information from Westinghouse. 

The request must state with particularity:

a. The nature of the proprietary or other information sought;

b. The reason why the information currently available to the public in the NRC’s public

document room is insufficient;

c. The relevance of the requested information to the hearing issue(s) which the person

proposes to raise; and

d. A showing that the requesting person has the capability to understand and utilize the

requested information. 

2. If a person claims that the information is necessary to prepare a request for hearing,

the request must be filed no later than 15 days after publication in the Federal Register of the

notice required either by 10 CFR 52.85 or 10 CFR 52.103. If Westinghouse declines to provide

the information sought, Westinghouse shall send a written response within ten (10) days of

receiving the request to the requesting person setting forth with particularity the reasons for its

refusal. The person may then request the Commission (or presiding officer, if a proceeding has

been established) to order disclosure. The person shall include copies of the original request

(and any subsequent clarifying information provided by the requesting party to the applicant)

and the applicant’s response. The Commission and presiding officer shall base their decisions

solely on the person’s original request (including any clarifying information provided by the

requesting person to Westinghouse), and Westinghouse’s response. The Commission and

presiding officer may order Westinghouse to provide access to some or all of the requested

information, subject to an appropriate non-disclosure agreement.
55

VII. DURATION OF THIS APPENDIX

This appendix may be referenced for a period of 15 years from [Insert date 30 days after

publication of the final rule in the Federal Register], except as provided for in 10 CFR 52.55(b)

and 52.57(b). This appendix remains valid for an applicant or licensee who references this

appendix until the application is withdrawn or the license expires, including any period of

extended operation under a renewed license. 

VIII. PROCESSES FOR CHANGES AND DEPARTURES

A. Tier 1 information. 

1. Generic changes to Tier 1 information are governed by the requirements in 10 CFR

52.63(a)(1). 

2. Generic changes to Tier 1 information are applicable to all applicants or licensees

who reference this appendix, except those for which the change has been rendered technically

irrelevant by action taken under paragraphs A.3 or A.4 of this section. 

3. Departures from Tier 1 information that are required by the Commission through

plant-specific orders are governed by the requirements in 10 CFR 52.63(a)(3). 

4. Exemptions from Tier 1 information are governed by the requirements in 10 CFR

52.63(b)(1) and § 52.97(b). The Commission will deny a request for an exemption from Tier 1,

if it finds that the design change will result in a significant decrease in the level of safety

otherwise provided by the design. 

B. Tier 2 information. 

1. Generic changes to Tier 2 information are governed by the requirements in 10 CFR

52.63(a)(1).

2. Generic changes to Tier 2 information are applicable to all applicants or licensees

who reference this appendix, except those for which the change has been rendered technically
56

irrelevant by action taken under paragraphs B.3, B.4, B.5, or B.6 of this section. 

3. The Commission may not require new requirements on Tier 2 information by

plant-specific order while this appendix is in effect under §§ 52.55 or 52.61, unless:

a. A modification is necessary to secure compliance with the Commission’s regulations

applicable and in effect at the time this appendix was approved, as set forth in Section V of this

appendix, or to ensure adequate protection of the public health and safety or the common

defense and security; and 

b. Special circumstances as defined in 10 CFR 50.12(a) are present. 

4. An applicant or licensee who references this appendix may request an exemption

from Tier 2 information. The Commission may grant such a request only if it determines that

the exemption will comply with the requirements of 10 CFR 50.12(a). The Commission will

deny a request for an exemption from Tier 2, if it finds that the design change will result in a

significant decrease in the level of safety otherwise provided by the design. The grant of an

exemption to an applicant must be subject to litigation in the same manner as other issues

material to the license hearing. The grant of an exemption to a licensee must be subject to an

opportunity for a hearing in the same manner as license amendments. 

5.a. An applicant or licensee who references this appendix may depart from Tier 2

information, without prior NRC approval, unless the proposed departure involves a change to or

departure from Tier 1 information, Tier 2* information, or the TS, or requires a license

amendment under paragraphs B.5.b or B.5.c of this section. When evaluating the proposed

departure, an applicant or licensee shall consider all matters described in the plant-specific

DCD. 

b. A proposed departure from Tier 2, other than one affecting resolution of a severe

accident issue identified in the plant-specific DCD, requires a license amendment if it would:

(1) Result in more than a minimal increase in the frequency of occurrence of an
57

accident previously evaluated in the plant-specific DCD;

(2) Result in more than a minimal increase in the likelihood of occurrence of a

malfunction of a structure, system, or component (SSC) important to safety and previously

evaluated in the plant-specific DCD;

(3) Result in more than a minimal increase in the consequences of an accident

previously evaluated in the plant-specific DCD;

(4) Result in more than a minimal increase in the consequences of a malfunction of an

SSC important to safety previously evaluated in the plant-specific DCD;

(5) Create a possibility for an accident of a different type than any evaluated previously

in the plant-specific DCD;

(6) Create a possibility for a malfunction of an SSC important to safety with a different

result than any evaluated previously in the plant-specific DCD;

(7) Result in a design basis limit for a fission product barrier as described in the plantspecific DCD being exceeded or altered; or

(8) Result in a departure from a method of evaluation described in the plant-specific

DCD used in establishing the design bases or in the safety analyses.

c. A proposed departure from Tier 2 affecting resolution of a severe accident issue

identified in the plant-specific DCD, requires a license amendment if –

(1) There is a substantial increase in the probability of a severe accident such that a

particular severe accident previously reviewed and determined to be not credible could become

credible; or 

(2) There is a substantial increase in the consequences to the public of a particular

severe accident previously reviewed.

d. If a departure requires a license amendment under paragraph B.5.b or B.5.c of this

section, it is governed by 10 CFR 50.90.
58

e. A departure from Tier 2 information that is made under paragraph B.5 of this section

does not require an exemption from this appendix.

f. A party to an adjudicatory proceeding for either the issuance, amendment, or renewal

of a license or for operation under 10 CFR 52.103(a), who believes that an applicant or licensee

who references this appendix has not complied with paragraph VIII.B.5 of this appendix when

departing from Tier 2 information, may petition to admit into the proceeding such a contention. 

In addition to compliance with the general requirements of 10 CFR 2.309, the petition must

demonstrate that the departure does not comply with paragraph VIII.B.5 of this appendix. 

Further, the petition must demonstrate that the change bears on an asserted noncompliance

with an ITAAC acceptance criterion in the case of a 10 CFR 52.103 preoperational hearing, or

that the change bears directly on the amendment request in the case of a hearing on a license

amendment. Any other party may file a response. If, on the basis of the petition and any

response, the presiding officer determines that a sufficient showing has been made, the

presiding officer shall certify the matter directly to the Commission for determination of the

admissibility of the contention. The Commission may admit such a contention if it determines

the petition raises a genuine issue of material fact regarding compliance with paragraph VIII.B.5

of this appendix. 

6.a. An applicant who references this appendix may not depart from Tier 2* information,

which is designated with italicized text or brackets and an asterisk in the generic DCD, without

NRC approval. The departure will not be considered a resolved issue, within the meaning of

Section VI of this appendix and 10 CFR 52.63(a)(4). 

b. A licensee who references this appendix may not depart from the following Tier 2*

matters without prior NRC approval. A request for a departure will be treated as a request for a

license amendment under 10 CFR 50.90.

(1) Maximum fuel rod average burn-up.
59

(2) Fuel principal design requirements.

(3) Fuel criteria evaluation process.

(4) Fire areas.

(5) Human factors engineering.

(6) Small-break loss-of-coolant (LOCA) Analysis Methodology.

c. A licensee who references this appendix may not, before the plant first achieves full

power following the finding required by 10 CFR 52.103(g), depart from the following Tier 2*

matters except under paragraph B.6.b of this section. After the plant first achieves full power,

the following Tier 2* matters revert to Tier 2 status and are subject to the departure provisions

in paragraph B.5 of this section. 

(1) Nuclear Island structural dimensions.

(2) American Society of Mechanical Engineers Boiler & Pressure Vessel Code (ASME

Code), Section III, and Code Case –284.

(3) Design Summary of Critical Sections.

(4) American Concrete Institute (ACI) 318, ACI 349, American National Standards

Institute/American Institute of Steel Construction (ANSI/AISC)–690, and American Iron and

Steel Institute (AISI).

(5) Definition of critical locations and thicknesses.

(6) Seismic qualification methods and standards.

(7) Nuclear design of fuel and reactivity control system, except burn-up limit. 

(8) Motor-operated and power-operated valves.

(9) Instrumentation and control system design processes, methods, and standards.

(10) Passive residual heat removal (PRHR) natural circulation test (first plant only).

(11) Automatic depressurization system (ADS) and core make-up tank (CMT)

verification tests (first three plants only).
60

(12) Polar Crane Parked Orientation.

(13) Piping design acceptance criteria.

(14) Containment Vessel Design Parameters.

d. Departures from Tier 2* information that are made under paragraph B.6 of this

section do not require an exemption from this appendix. 

C. Operational requirements.

1. Generic changes to generic TS and other operational requirements that were

completely reviewed and approved in the design certification rulemaking and do not require a

change to a design feature in the generic DCD are governed by the requirements in 10 CFR

50.109. Generic changes that do require a change to a design feature in the generic DCD are

governed by the requirements in paragraphs A or B of this section.

2. Generic changes to generic TS and other operational requirements are applicable to

all applicants or licensees who reference this appendix, except those for which the change has

been rendered technically irrelevant by action taken under paragraphs C.3 or C.4 of this

section. 

3. The Commission may require plant-specific departures on generic TS and other

operational requirements that were completely reviewed and approved, provided a change to a

design feature in the generic DCD is not required and special circumstances as defined in

10 CFR 2.335 are present. The Commission may modify or supplement generic TS and other

operational requirements that were not completely reviewed and approved or require additional

TS and other operational requirements on a plant-specific basis, provided a change to a design

feature in the generic DCD is not required. 

4. An applicant who references this appendix may request an exemption from the

generic TS or other operational requirements. The Commission may grant such a request only

if it determines that the exemption will comply with the requirements of 10 CFR 50.12(a). The
61

grant of an exemption must be subject to litigation in the same manner as other issues material

to the license hearing. 

5. A party to an adjudicatory proceeding for either the issuance, amendment, or

renewal of a license or for operation under 10 CFR 52.103(a), who believes that an operational

requirement approved in the DCD or a TS derived from the generic TS must be changed may

petition to admit such a contention into the proceeding. The petition must comply with the

general requirements of 10 CFR 2.309 and must demonstrate why special circumstances as

defined in 10 CFR 2.335 are present, or demonstrate compliance with the Commission’s

regulations in effect at the time this appendix was approved, as set forth in Section V of this

appendix. Any other party may file a response to the petition. If, on the basis of the petition

and any response, the presiding officer determines that a sufficient showing has been made,

the presiding officer shall certify the matter directly to the Commission for determination of the

admissibility of the contention. All other issues with respect to the plant-specific TS or other

operational requirements are subject to a hearing as part of the license proceeding. 

6. After issuance of a license, the generic TS have no further effect on the plantspecific TS. Changes to the plant-specific TS will be treated as license amendments under

10 CFR 50.90. 

IX. INSPECTIONS, TESTS, ANALYSES, AND ACCEPTANCE CRITERIA (ITAAC)

A.1 An applicant or licensee who references this appendix shall perform and

demonstrate conformance with the ITAAC before fuel load. With respect to activities subject to

an ITAAC, an applicant for a license may proceed at its own risk with design and procurement

activities. A licensee may also proceed at its own risk with design, procurement, construction,

and preoperational activities, even though the NRC may not have found that any particular

ITAAC has been satisfied.
62

2. The licensee who references this appendix shall notify the NRC that the required

inspections, tests, and analyses in the ITAAC have been successfully completed and that the

corresponding acceptance criteria have been met.

3. If an activity is subject to an ITAAC and the applicant or licensee who references this

appendix has not demonstrated that the ITAAC has been satisfied, the applicant or licensee

may either take corrective actions to successfully complete that ITAAC, request an exemption

from the ITAAC under Section VIII of this appendix and 10 CFR 52.97(b), or petition for

rulemaking to amend this appendix by changing the requirements of the ITAAC, under 10 CFR

2.802 and 52.97(b). Such rulemaking changes to the ITAAC must meet the requirements of

paragraph VIII.A.1 of this appendix.

B.1 The NRC shall ensure that the required inspections, tests, and analyses in the

ITAAC are performed. The NRC shall verify that the inspections, tests, and analyses

referenced by the licensee have been successfully completed and find that the prescribed

acceptance criteria have been met. At appropriate intervals during construction, the NRC shall

publish notices of the successful completion of ITAAC in the Federal Register.

2. Under 10 CFR 52.99 and 52.103(g), the Commission shall find that the acceptance

criteria in the ITAAC for the license are met before fuel load. 

3. After the Commission has made the finding required by 10 CFR 52.103(g), the

ITAAC do not, by virtue of their inclusion within the DCD, constitute regulatory requirements

either for licensees or for renewal of the license; except for specific ITAAC, which are the

subject of a Section 103(a) hearing, their expiration will occur upon final Commission action in

such a proceeding. However, subsequent modifications must comply with the Tier 1 and Tier 2

design descriptions in the plant-specific DCD unless the licensee has complied with the

applicable requirements of 10 CFR 52.97 and Section VIII of this appendix.
63

X. RECORDS AND REPORTING

A. Records

1. The applicant for this appendix shall maintain a copy of the generic DCD that

includes all generic changes to Tier 1 and Tier 2. The applicant shall maintain the proprietary

and safeguards information referenced in the generic DCD for the period that this appendix

may be referenced, as specified in Section VII of this appendix. 

2. An applicant or licensee who references this appendix shall maintain the

plant-specific DCD to accurately reflect both generic changes to the generic DCD and

plant-specific departures made under Section VIII of this appendix throughout the period of

application and for the term of the license (including any period of renewal). 

3. An applicant or licensee who references this appendix shall prepare and maintain

written evaluations which provide the bases for the determinations required by Section VIII of

this appendix. These evaluations must be retained throughout the period of application and for

the term of the license (including any period of renewal). 

B. Reporting 

1. An applicant or licensee who references this appendix shall submit a report to the

NRC containing a brief description of any departures from the plant-specific DCD, including a

summary of the evaluation of each. This report must be filed in accordance with the filing

requirements applicable to reports in 10 CFR 50.4. 

2. An applicant or licensee who references this appendix shall submit updates to its

DCD, which reflect the generic changes to and plant-specific departures from the generic DCD

made under Section VIII of this appendix. These updates shall be filed under the filing

requirements applicable to final safety analysis report updates in 10 CFR 50.4 and 50.71(e). 

3. The reports and updates required by paragraphs X.B.1 and X.B.2 must be submitted

as follows:
64

a. On the date that an application for a license referencing this appendix is submitted,

the application shall include the report and any updates to the generic DCD.

b. During the interval from the date of application for a license to the date the

Commission makes its findings under 10 CFR 52.103(g), the report must be submitted

semi-annually. Updates to the plant-specific DCD must be submitted annually and may be

submitted along with amendments to the application. 

c. After the Commission has made its finding under 10 CFR 52.103(g), the reports and

updates to the plant-specific DCD must be submitted, along with updates to the site-specific

portion of the final safety analysis report for the facility, at the intervals required by 10 CFR

50.59(d)(2) and 50.71(e)(4), respectively, or at shorter intervals as specified in the license. 

Dated at Rockville, Maryland, this day of 2004.

For the Nuclear Regulatory Commission.

_____________________________________

Annette L. Vietti-Cook

Secretary of the Commission
ENVIRONMENTAL ASSESSMENT BY THE

U.S. NUCLEAR REGULATORY COMMISSION

RELATING TO THE CERTIFICATION OF THE 

AP1000 STANDARD PLANT DESIGN

DOCKET NO. 52-006
-2-

Table of Contents

1.0 Identification of the Proposed Action .......................................4

2.0 The Need for the Proposed Action ........................................4

3.0 The Environmental Impact of the Proposed Action ............................5

4.0 Alternatives to the Proposed Action .......................................6

4.1 Severe Accident Mitigation Design Alternatives ...........................7

4.2 Potential SAMDAs Identified by Westinghouse . . . . . . . . . . . . . . . . . . . . . . . . . . 10

4.3 NRC Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

4.4 Risk Reduction Potential of SAMDAs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

4.4.1 Westinghouse Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

4.4.2 NRC Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

4.5 Cost Impacts of Candidate SAMDAs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

4.5.1 Westinghouse Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20

4.5.2 NRC Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

4.6 Cost-Benefit Comparison . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

4.6.1 Westinghouse Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21

4.6.2 NRC Evaluation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23

4.7 Further Considerations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25

4.7.1 Self-Actuating Containment Isolation Valves . . . . . . . . . . . . . . . . . . . . . 25

4.7.2 Diverse IRWST Injection Valves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26

4.8 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27

5.0 Alternative Use of Resources . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28

6.0 States Consulted and Sources Used . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28
-3-

UNITED STATES NUCLEAR REGULATORY COMMISSION

ENVIRONMENTAL ASSESSMENT AND FINDING OF

NO SIGNIFICANT IMPACT

RELATING TO THE CERTIFICATION OF THE 

AP1000 STANDARD PLANT DESIGN

DOCKET NO. 52-006

The U.S. Nuclear Regulatory Commission (NRC) has issued a design certification for

the Advanced Passive 1000 (AP1000) design in response to an application submitted on

March 28, 2002, by Westinghouse Electric Company LLC (hereinafter referred to as

Westinghouse). A design certification is a rulemaking that amends 10 CFR Part 52. 

The NRC has performed an environmental assessment (EA) of the environmental

impacts of the proposed new rule and has documented its findings of no significant impact in

accordance with the requirements of 10 CFR 51.21 and the National Environmental Policy Act

of 1969 (NEPA), as amended. This EA also addresses the severe accident mitigation design

alternatives (SAMDAs), that the NRC has to considered as part of this EA for the AP1000

design. This EA does not address the site-specific environmental impacts of constructing and

operating a facility, that references the AP1000 design certification; such impacts will be

evaluated as part of any application or applications for the siting, construction, or operation of a

facility.

As discussed in detail in Section 4.0 of this EA, the NRC determined that issuing this

design certification does not constitute a major Federal action significantly affecting the quality

of the human environment. The basis for this finding of no significant impact is that the design

certification would not authorize the siting, construction, or operation of a facility of an AP1000

reactor design. Rather, the certification would merely codify the AP1000 design in a rule that
-4-

could be referenced in a construction permit (CP), combined license (COL), or operating license

(OL) application. Further, because the certification is just a rule, it does not involve any

resources that have alternative uses. Therefore, the NRC has not prepared an environmental

impact statement (EIS) in connection with this action. 

The NRC also reviewed Westinghouse’s evaluation of SAMDAs that generically apply to

the AP1000 design. On that basis, the NRC found that the evaluation provides reasonable

assurance that there are no additional SAMDAs beyond those currently incorporated into the

AP1000 design which are cost-beneficial, whether considered at the time of the approval of the

AP1000 design certification or in connection with the licensing of a future facility referencing the

AP1000 design certification, where the plant referencing this appendix is located on a site

whose site parameters are within those specified in Appendix 1B of the AP1000 design control

document (DCD). These issues are considered resolved for the AP1000 design.

ENVIRONMENTAL ASSESSMENT

1.0 Identification of the Proposed Action

The proposed action would certify the AP1000 design under Appendix D to

10 CFR Part 52. The new rule would allow prospective licensees to reference the certified

AP1000 design as part of a combined license (COL) application under 10 CFR Part 52 or may

allow for a construction permit (CP) application under 10 CFR Part 50. 

2.0 The Need for the Proposed Action

The NRC has long sought the safety benefits of commercial nuclear power plant

standardization, and early final resolution of design issues. The NRC plans to achieve these

benefits by certifying nuclear plant designs. Subpart B to 10 CFR Part 52 allows for certification

in the form of rulemaking of an essentially complete nuclear plant design. 
-5-

The proposed action would amend 10 CFR Part 52 to certify the AP1000 design. The

amendment would allow prospective licensees to reference the certified AP1000 design as part

of a COL application under 10 CFR Part 52 or may allow for a CP application under

10 CFR Part 50. Those portions of the AP1000 design included in the scope of the certification

rulemaking would not be subject to further safety regulatory review or approval in a COL

proceeding. In addition, the design certification rule would eliminate the need to consider

SAMDAs for any future facilities that reference the certified AP1000 design. 

3.0 The Environmental Impact of the Proposed Action

Issuing an amendment to 10 CFR Part 52 to certify the AP1000 standard plant design

would not constitute a significant environmental impact. The amendment would merely codify

the NRC’s approval of the AP1000 design (refer to NUREG-1793). Furthermore, because the

amendment is a rule, it involves no resources that have alternative uses. 

As described in Section 4.0 of this EA, the NRC reviewed alternatives to the design

certification rulemaking and alternative design features for preventing and mitigating severe

accidents. NEPA requires consideration of alternatives to show that the design certification rule

is the appropriate course of action and to ensure that the design referenced in the rulemaking

does not exclude any cost-beneficial design changes related to the prevention and mitigation of

severe accidents. The NRC concludes that, unlike the proposed design certification rule, the

alternatives to certification do not provide for resolution of issues. 

Design certification is in keeping with the Commission’s intent to make future plants

safer than the current generation of plants, to achieve early resolution of licensing issues, and

to achieve the safety benefits of standardization (refer to the Advanced Reactor (51 FR 24643),

Standardization (52 FR 348803), and Severe Accident Policy Statements (50 FR 32138), and to

10 CFR Part 52). Through its own independent analysis, the NRC also concludes that
-6-

Westinghouse adequately considered an appropriate set of SAMDAs and that none were costbeneficial. Although Westinghouse made no design changes as a result of reviewing the

SAMDAs, Westinghouse had already incorporated certain features in the AP1000 design on the

basis of the probabilistic risk assessment (PRA) results. Section 4.2 of this EA gives examples

of these features. These design features relate to severe accident prevention and mitigation,

but were not considered in the SAMDA evaluation because they were already part of the

AP1000 design (refer to Section 19.1.6.2 of NUREG-1793, ?AP1000 Design Improvement as a

Result of Probabilistic Risk Assessment Studies”).

Finally, the design certification rule by itself would not authorize the siting, construction, 

or operation of a nuclear power plant. The issuance of a CP, early site permit (ESP), COL, or

OL which references the AP1000 design will require a prospective applicant to address the

environmental impacts of construction and operation at a specific site. The NRC will then

evaluate the environmental impacts and issue an EIS in accordance with 10 CFR Part 51. 

However, the SAMDA analysis has been completed as part of this EA and can be incorporated

by reference into an EIS related to siting, construction, or operation of a nuclear plant that

references the AP1000 design. 

4.0 Alternatives to the Proposed Action

The NRC has two alternatives to certifying the AP1000 design in an amendment to 

10 CFR Part 52: take no action to approve the design or issue a final design approval (FDA)

without certifying the design. As with the proposed action, these alternatives would not have a

significant impact on the quality of the human environment because they would not authorize

the siting, construction, or operation of a facility. 

In the first case, the NRC would not approve the design. Therefore, as a custom plant,

a facility built using the AP1000 design would require licensing under 10 CFR Part 50 or 
-7-

10 CFR Part 52, Subpart C. Moreover, all design issues would have to be considered as part

of each application to construct and operate an AP1000 design at a particular site. As a result,

this alternative would neither achieve the benefits of standardization, provide for early resolution

of design issues, nor permit finality for the resolved design issues. 

In the second case, the NRC would issue an FDA under Appendix O to 10 CFR Part 52,

but would not certify the design in a rulemaking. Therefore, although the NRC would have

approved the design, the design could be modified and, thus, would require re-evaluation as

part of each application to construct and operate a facility of an AP1000 design at a particular

site. This alternative would provide for early internal NRC resolution of design issues to the

extent that the design would remain unchanged at the facility application stage, but would not

obtain the benefits of standardization nor permit overall finality for the resolved design issues. 

The NRC sees no advantage in either alternative compared to the design certification

rulemaking proposed for the AP1000 design. Although neither the alternatives nor the

proposed action (design certification rulemaking) would significantly affect the quality of the

human environment, the proposed action achieves the benefits of standardization, permits early

resolution of design issues, and provides finality for the resolved design issues (including

SAMDAs) that are within the scope of the design certification. Therefore, the NRC concludes

that the alternatives to rulemaking would not achieve the objectives that the Commission

intends by certifying the AP1000 design pursuant to 10 CFR Part 52, Subpart B. 

4.1 Severe Accident Mitigation Design Alternatives

Consistent with the objectives of standardization and early resolution of design issues,

the Commission decided to evaluate SAMDAs as part of the design certification for the AP1000

design. In a 1985 policy statement, the Commission defined the term ?severe accident” as an

event that is ?beyond the substantial coverage of design-basis events,” including events where
 4

Although 10 CFR 50.34(f)(1)(I) by its terms does not apply to new construction permits (CP), the

Commission’s policy is that a CP applicant will be required to comply with 50.34(f)(1)(I).

-8-

there is substantial damage to the reactor core (whether or not there are serious offsite

consequences). Design-basis events are events analyzed in accordance with the NRC’s

Standard Review Plan (NUREG-0800) and documented in Chapter 15 of the DCD.

As part of its design certification application, Westinghouse performed a PRA for the

AP1000 design to achieve the following objectives:

C Identify the dominant severe accident sequences and associated source terms for the

design. 

C Modify the design, on the basis of PRA insights, to prevent or mitigate and reduce the

risk of severe accidents. 

C Provide a basis for concluding that all reasonable steps have been taken to reduce the

chances of occurrence, and mitigate the consequences, of severe accidents. 

Westinghouse’s PRA analysis is described in Chapter 19 of the AP1000 DCD.

In addition to considering alternatives to the rulemaking process discussed in

Section 3.0, applicants for reactor design certification, COLs, or CPs must also consider

alternative design features for severe accidents consistent with the requirements of 

10 CFR Part 50, and with a court ruling related to NEPA. These requirements can be

summarized as follows:

C 10 CFR 52.79, and 10 CFR 50.34(f)(1)(I)4

 requires the applicant to perform a

plant/site-specific PRA, the aim of which is to seek such improvements in the reliability

of core and containment heat removal systems as are significant and practical and do

not impact excessively on the plant.
-9-

C The U.S. Court of Appeals decision, in Limerick Ecology Action v. NRC, 869 F.2d 719

(3rd Cir. 1989), effectively requires the NRC to consider certain SAMDAs in the

environmental impact review performed under Section 102(2)(c) of NEPA.

Although these requirements are not directly related, they share a common purpose to

consider alternatives to the proposed design, to evaluate whether potential alternative

improvements in the plant design might increase safety performance during severe accidents,

and to prevent reasonable alternatives from being foreclosed. It should be noted that the

Commission is not required to consider alternatives to the design in this EA. However, as a

matter of discretion, the Commission has determined that considering SAMDAs concomitant

with the rulemaking is consistent with the intent of 10 CFR Part 52 for early resolution of issues,

finality for resolved design issues, and achieving the benefits of standardization.

In its decision in Limerick Ecology Action v. NRC, the Court of Appeals for the Third

Circuit expressed its opinion that it would likely be difficult to evaluate SAMDAs for NEPA

purposes on a generic basis. However, the NRC has determined that generic evaluation of

SAMDAs for the AP1000 standard design is warranted for two significant reasons. First, the

design and construction of all plants referencing the certified AP1000 design will be governed

by the rule certifying a single design. Second, the site parameters specified in the rule and the

AP1000 DCD establish the consequences for a reasonable enveloping set of SAMDAs for the

AP1000 design. The low residual risk of the AP1000 design and the limited potential for further

risk reductions provides high confidence that additional cost-beneficial SAMDAs would not be

found for sites within the site parameter envelope assumed for the AP1000 EA of SAMDAs. If

the actual parameters for a particular site exceed those assumed in the rule and the DCD, then

SAMDAs must be re-evaluated in the site-specific environmental report and the EIS. If the

actual parameters for a postulated site are bounded by those assumed in the rule and the DCD,

then the SAMDA analysis can be incorporated by reference in the site-specific EIS. 
-10-

4.2 Potential SAMDAs Identified by Westinghouse

To identify candidate design alternatives, Westinghouse reviewed the design

alternatives for other plants including the CE System 80+. Westinghouse also reviewed the

results of the AP1000 PRA and design alternatives suggested by AP1000 design personnel. 

Westinghouse eliminated the following SAMDAs from further consideration because 

they are already incorporated in the AP1000 design:

C hydrogen ignition system

C reactor cavity flooding system

C reactor coolant pump seal cooling (AP1000 has canned motor pumps)

C reactor coolant system (RCS) depressurization

C external reactor vessel cooling

C non-safety-grade containment sprays

Several risk-significant enhancements to the AP600 design have also been incorporated

in the AP1000 design and were therefore not further considered. These modifications are

summarized below and discussed further in DCD Tier 2, Section 1B.1.5, “Summary of Risk

Significant Enhancement.”

• a change in the normal position of the two containment motor-operated recirculation 

valves (in series with squib valves) from closed to open to improve the reliability of

opening these flowpaths

• a change in the emergency operating procedures (EOPs) to call for in-containment

refueling water storage tank (IRWST) draining earlier in an event to improve the

probability of successful operator action

• a change in the design of the IRWST vents to preferentially direct hydrogen releases to 

the IRWST pipe vents, where diffusion flames will not adversely impact the containment
-11-

• incorporation of a low-boron core to reduce the potential contribution of anticipated

transient without scram (ATWS) events to plant risk 

• addition of a third passive containment cooling system (PCS) drainline with a

motor-operated valve (MOV) that is diverse from the air-operated valves (AOVs) used in

the other two drainlines, to improve PCS reliability

• specification that two of the four squib valves in the recirculation lines be low-pressuretype valves, and the remaining two squib valves be high-pressure type valves to reduce

the contribution to core damage frequency (CDF) from common-cause failures (CCFs)

of recirculation squib valves

On the basis of the screening, Westinghouse retained 14 potential SAMDAs for further

consideration. This set of SAMDAs is the same as that considered for the AP600 design. DCD

Tier 2, Section 1B.1.3, “Selection and Description of SAMDAs,” describes the 14 design

improvements as follows:

1. Upgrade the chemical and volume control system (CVCS) for small loss-of-coolant

accidents (LOCAs): The CVCS is currently capable of maintaining the RCS inventory

for LOCAs for effective break sizes up to 0.97 cm (3/8 in.) in diameter. A design

alternative involving the upgrade of the CVCS for small LOCAs would increase the

capability of the CVCS, enabling it to maintain RCS inventory during small- and

intermediate-size LOCAs (up to an effective break size of 15.2 cm (6 in.) in diameter). 

Implementation of this design alternative would require installation of IRWST and

containment recirculation connections to the CVCS, as well as the addition of a second

line from the CVCS pumps to the RCS.

2. Filtered vent: This design alternative would involve the installation of a filtered

containment vent, including all associated piping and penetrations. This modification
-12-

would provide a means to vent containment to prevent catastrophic overpressure

failures and would also provide a filtering capability for source term release. The filtered

vent would reduce the risk of late containment failures that might occur after failure of

the PCS. Note, however, that even if the PCS fails, it is expected that air cooling will

limit the containment pressure to less than the ultimate pressure capacity of the

containment under most environmental conditions.

3. Self-actuating containment isolation valves (CIVs): Self-actuation of CIVs could be used

to increase the likelihood of successful containment isolation during a severe accident. 

This design alternative would involve the addition of a self-actuating valve or the

enhancement of the existing CIVs on normally open containment penetrations

(i.e., penetrations that provide normally open pathways to the environment during power

and normal shutdown conditions). The design alternative would provide for

self-actuation in the event that containment conditions are indicative of a severe

accident. Closed systems inside and outside containment, such as the normal residual

heat removal system (RHRS) and component cooling, would be excluded from this

design alternative. The actuation of CIVs would be automatically initiated in the event

that containment conditions are indicative of a severe accident.

4. Passive containment sprays: This SAMDA involves adding a passive safety-related

spray system and all associated piping and support systems to the AP1000 design

(in lieu of the non-safety-related active containment spray capability currently

incorporated in the AP1000 design). Installation of the safety-grade containment spray

system could result in an increase in the following three risk benefits:

1. scrubbing of fission products, primarily for containment isolation failure

• alternative means for flooding the reactor vessel (in-vessel retention)

• control of containment pressure if the PCS fails
-13-

5. Active high pressure safety injection system (HPSI): A safety-related active HPSI

system could be added that would be capable of preventing a core melt for all events

except the large-break LOCA and ATWS. Note, however, that this design alternative is

not consistent with the AP1000 design objectives. The AP1000 would change from a

plant with passive systems to a plant with passive and active systems.

6. Steam generator (SG) shell-side heat removal system: This design alternative would

involve the installation of a passive safety-related heat removal system to the secondary

side of the SGs. This enhancement would provide closed-loop secondary-system

cooling by means of natural circulation and stored water cooling, thereby preventing the

loss of the primary heat sink given the loss of startup feedwater (SFW) and the passive

RHR heat exchanger (HX).

7. Direct steam generator relief flow to the IRWST: To prevent fission product release

from bypassing containment during an steam generator tube rupture (SGTR) event (or

to reduce the amount released), flow from the SG safety and relief valves could be

directed to the IRWST. An alternative, lower cost option would be to redirect flow only

from the first-stage safety valve to the IRWST.

8. Increased steam generator pressure capability: As an alternative to design

alternative (7) above, another method could be used to prevent fission product release

from bypassing containment during an SGTR event (or to reduce the amount). This

alternative method would involve an increase of the SG secondary-side and safety valve

setpoint to a level high enough to not allow an SGTR to cause the secondary-system

safety valve to open. Although detailed analyses have not been performed, it is

estimated that the secondary-side design pressure would have to be increased by

several hundred pounds per square inch (psi).
-14-

9. Secondary containment filtered ventilation: This design alternative involves the

installation of a passive charcoal and high-efficiency particulate air filter system for the

middle- and lower-annulus region of the secondary concrete containment (below

elevation 135'-3"). Drawing a partial vacuum on the middle annulus via an eductor with

motive power from compressed gas tanks would operate the filter system. This design

alternative would reduce particulate fission product release from any failed containment

penetrations.

10. Diverse IRWST injection valves: In the current design, a squib valve in series with a

check valve (CV) isolates each of the four IRWST injection paths. To provide diversity,

a modification could be made to allow a different vendor to provide the valves in two of

the lines. Such diverse IRWST injection valves would reduce the likelihood of CCFs of

the four IRWST injection paths.

11. Diverse containment recirculation valves: In both the AP600 and AP1000 designs, two

of the four recirculation lines have a squib valve in series with a CV, and the remaining

two recirculation lines have a squib valve in series with an MOV. This SAMDA involves

changing the recirculation valve specification to enable two of the four lines to use

diverse squib valves. To provide diversity, a modification could be made to allow a

different vendor to provide the squib valves in two lines. Alternatively, in the AP1000

design, Westinghouse has specified that two of the four recirculation squib valves be

designated as the low-pressure type and the remaining two squib valves as the high

pressure type. The diverse containment recirculation valves incorporated in the AP1000

design are responsive to the intent of this SAMDA and will reduce the frequency of core

melt due to CCF of the four containment recirculation lines.
-15-

12. Ex-vessel core catcher: This design alternative would inhibit core concrete interaction

(CCI), even if the debris bed dries out. The enhancement would involve the design of a

structure in the containment cavity or the use of a special concrete or coating. The

current AP1000 design incorporates a wet cavity design in which ex-vessel cooling is

used to keep core debris within the vessel. In cases where reactor vessel flooding has

failed, the PRA assumes that containment failure occurs from an ex-vessel steam

explosion or CCI.

13. High-pressure containment design: A high-pressure containment design would prevent

containment failures from severe accident phenomena such as steam explosions and

hydrogen detonation. This proposed containment design would have a design pressure

of approximately 2.17 MPa (300 psig) and would include a passive cooling feature

similar to the one in the existing containment design. Although the high-pressure

containment would not reduce the frequency or magnitude of releases from an

unisolated containment, it would reduce the likelihood of containment failures.

14. Increase reliability of diverse actuation system (DAS): The DAS is a non-safety system

that can automatically trip the reactor and turbine and actuate certain engineered safety

feature (ESF) equipment if the protection and safety monitoring system (PMS) is unable

to perform these functions. The DAS provides diverse monitoring of selected plant

parameters to guide manual operation and to confirm reactor trip and ESF actuations. 

Increasing the reliability of the DAS involves adding a third instrumentation and control

(I&C) cabinet and a third set of DAS instruments to allow the use of two-out-of-three

logic instead of two-out-of-two logic.

Westinghouse considered an additional SAMDA that would involve relocating the entire

normal residual heat removal system (RNS) and piping inside the containment pressure

boundary. This would prevent containment bypass due to intersystem loss-of-coolant accidents
-16-

(ISLOCAs) in the RNS. However, in the AP1000, the RNS has a higher design pressure than

the systems in current pressurized-water reactors (PWRs), and an additional isolation valve is

provided. As a result, ISLOCAs do not contribute significantly to the CDF in the AP1000 PRA. 

Accordingly, Westinghouse did not further investigate this change. The NRC has reviewed the

Westinghouse analyses and agrees that further consideration of this change is not warranted

because the change would provide virtually no risk reduction.

4.3 NRC Evaluation

The set of potential design improvements considered for the AP1000 is the same as

those considered for the AP600. As part of the review for the AP600, the NRC reviewed the set

of potential design improvements identified by Westinghouse and found it to be reasonably

complete. The activity was accomplished by reviewing design alternatives associated with the

following plants: Limerick, Comanche Peak, CE System 80+, Watts Bar, and the advanced

boiling water reactor (ABWR). The NRC also reviewed accident management strategies

described in (NUREG/CR-5474) and alternatives identified through the Containment

Performance Improvement (CPI) Program (NUREG/CR-5567, -5575, -5630, and -5562). The

results of this assessment are summarized in Appendix A to “Review of Severe Accident

Mitigation Design Alternatives (SAMDAs) for the Westinghouse AP600 Design,” Science and

Engineering Associates, Inc., (SEA 97-2708-010-A;1, August 29, 1997). Given the similarity

between the AP1000 and the AP600 design features and risk profile, the NRC considers this

prior evaluation for the AP600 to be applicable to the AP1000 as well.

The NRC notes that the AP1000 design is less tolerant of equipment failures than the

AP600 because the large-LOCA success criterion for the AP1000 requires operation of two of

two accumulators whereas only one of two accumulators is required for the AP600, and

because the LOCA success criterion for the AP1000 requires operation of three of four

automatic depressurization system (ADS) Stage 4 valves whereas only two of four ADS Stage 4
-17-

valves are required for the AP600. At the NRC’s request, Westinghouse performed an

evaluation of the two additional design alternatives:

(1) Larger accumulators: An increase in the size of the accumulators sufficient to change

the large LOCA success criterion from two of two accumulators to one of two

accumulators. Westinghouse estimates that the accumulator tanks would have to

increase in size from 56.6 m3 to 113.2 m3 (2000 ft3 to 4000 ft3

). This increase would

likely require a change to the design of the direct vessel injection (DVI) piping

subsystem and significant reanalysis of the DVI piping.

(2) Larger ADS Stage 4 valves: Increasing the size of the ADS Stage 4 (ADS-4) valves

sufficient to change the LOCA success criterion from three of four valves to two of four

valves. Westinghouse estimates that the valves would have to increase in size from

35.6 cm to 45.7 cm (14 in. to 18 in.) and that common fourth stage piping that connects

to the hot leg would have to increase in size from 45.7 cm to 50.8 cm (18 in. to at least

20 in.). This increase would require a significant redesign of the squib valve and the

ADS-4 piping, which in turn would impact the design of the reactor coolant loop piping. 

Such a redesign would necessitate additional confirmatory testing to verify that the

behavior of the passive safety systems was not adversely impacted.

For both of these alternatives, Westinghouse estimated that the redesign and reanalysis

cost of the changes would be significantly greater than the benefits of completely eliminating all

severe accident risk for the AP1000. Therefore, these design changes were not pursued

further.

Although Westinghouse’s analysis omitted several design alternatives, in most

instances these design alternatives are either already included in the AP1000 design or

bounded in terms of risk reduction by one or more of the design alternatives that were included
-18-

in Westinghouse’s analysis. In some other cases, design alternatives were pertinent only to

boiling-water reactors (BWRs). The NRC’s review did not reveal any obvious additional design

alternatives that should have been considered by Westinghouse. Westinghouse considered

some of the potential design alternatives identified in the above references as appropriate for

accident management strategies, rather than as design alternatives. The NRC notes that the

set of design improvements is not all inclusive in that additional, perhaps less expensive design

improvements could be postulated. However, the benefits of any additional modifications would

not likely exceed the costs of the modifications evaluated. Also, the costs of alternative

improvements are not expected to be less than the costs of the least expensive improvements

evaluated, when the subsidiary costs associated with maintenance, procedures, and training

are considered.

The discussions in DCD Tier 2, Appendix 1B, do not provide Westinghouse’s basis or

process for screening the many possible design alternatives to arrive at the final list of 14. 

Although the information provided does not demonstrate that the search for design alternatives

was comprehensive, the NRC’s review of the more than 120 candidate design alternatives

considered for the AP600 did not identify any new alternatives more likely to be cost-beneficial

than those included in the AP1000 design alternative evaluations. The NRC notes that

Westinghouse has incorporated several risk significant enhancements within the AP1000

design, as discussed in Section 19.4.3.1 of NUREG-1793, ?Final Safety Evaluation Report

[FSER] Related to Certification of the AP1000 Standard Design,” (AP1000 FSER), and has

considered potential design changes to improve the AP1000 success criteria. On this basis,

the NRC concludes that the set of potential design improvements evaluated by Westinghouse

is acceptable.
-19-

4.4 Risk Reduction Potential of SAMDAs

4.4.1 Westinghouse Evaluation

In its evaluation, Westinghouse assumed that each design alternative would work

perfectly to completely eliminate all severe accident risk from evaluated internal, external, and

shutdown events. This assumption is conservative, since it maximizes the benefit of each

design alternative. The design alternative benefits were estimated on the basis of the reduction

of risk expressed in terms of whole body person-rem per year received by the total population

within a 80.5-km (50-mile) radius of the AP1000 plant site, as discussed in Section 19.4.2 of the

AP1000 FSER.

Westinghouse used the cost-benefit methodology of NUREG/BR-0184 to calculate the

maximum attainable benefit of completely eliminating all risk for the AP1000. This methodology

includes consideration of replacement power costs. The applicant estimated the present worth

of eliminating all risk to be $21,000. Even if the AP1000 CDF and large release frequency

(LRF) were a factor of 10 higher, this value would increase to only about $200,000.

4.4.2 NRC Evaluation

NRC reviewed Westinghouse’s bases for estimating the risk reduction for the various

SAMDAs, and concluded that Westinghouse used bounding and conservative assumptions as

the bases for the risk reduction estimates for each design alternative. 

Westinghouse’s risk reduction estimates are based on point-estimate (mean) values,

and do not consider uncertainties in CDF or offsite consequences. Although this is consistent

with the approach taken in previous design alternative evaluations, further consideration of

these factors could lead to significantly higher risk reduction values, given the extremely small

CDF and risk estimates in the baseline PRA. In assessing the risk reduction potential of design

improvements for the AP1000, the NRC has based its evaluation on the applicant’s risk
-20-

reduction estimates for the various design alternatives, in conjunction with an assessment of

the potential impact of uncertainties on the results. This assessment is discussed further in

Section 19.4.6 of the AP1000 FSER and in Section 4.6 of this EA. 

4.5 Cost Impacts of Candidate SAMDAs

4.5.1 Westinghouse Evaluation

DCD Tier 2, Section 1B.1.8, “Evaluation of Potential Improvements,” discusses capital

cost estimates for the design alternatives evaluated by Westinghouse for the AP1000. DCD

Tier 2, Table 1B-5, presents the results of the cost evaluations. The cost evaluations did not

account for the costs of design engineering, testing, and maintenance for each design

alternative. Including these costs would increase the overall costs and decrease the benefits of

each alternative. Thus, the Westinghouse approach is conservative.

4.5.2 NRC Evaluation

As mentioned previously, the set of SAMDAs considered for the AP1000 is the same as

the set considered for the AP600. As part of the AP600 review, the NRC compared the capital

costs for the AP600 design alternatives with those evaluated for the ABWR and CE System 80+

designs. The purpose of this comparison was to determine the reasonableness of the cost

estimates presented by the applicant. The design alternatives among the reactor designs, did

not exactly match, so only rough comparisons were possible. Based on these comparisons, the

NRC concluded that the cost estimates for the AP600 design alternatives are in reasonable

agreement with the costs for roughly similar design alternatives evaluated for other plants. 

Given the similarity between the AP1000 and the AP600 design features and risk profile, the

NRC considers this prior evaluation for the AP600 to be applicable to the AP1000 as well. This

is reasonable, considering uncertainties in the cost estimates, and the level of precision
-21-

necessary, given the greater uncertainty inherent on the benefit side with which these costs

were compared.

4.6 Cost-Benefit Comparison

4.6.1 Westinghouse Evaluation

After considering the risk reduction potential and cost impact of the various SAMDAs,

Westinghouse did a cost-benefit comparison to determine whether any of the potential severe

accident design features would be justified. To do so, Westinghouse evaluated the benefits of

each design alternative in terms of potential risk reduction, which was defined as the reduction

in whole body person-rem per year received by the total population within a 80.5-km (50-mile)

radius of the AP1000 plant site. Westinghouse used the cost-benefit methodology of 

NUREG/BR-0184 to calculate the maximum attainable benefit of completely eliminating all risk

for the AP1000. This methodology includes consideration of replacement power costs. 

Westinghouse estimated the present worth of eliminating all risk to be $21,000. This value is

an upper bound because in practice no design alternative, if implemented, would reduce the

plant CDF to zero. Westinghouse also provided additional sensitivity analyses of the impacts of

the following:

• a 3-percent discount rate rather than the 7-percent discount rate assumed in the base

case

IV. a factor of 10 increase in the population dose used in the base case

• a more realistic reduction in CDF (i.e., each SAMDA reduces CDF by 50 percent rather

than 100 percent, as assumed in the base case)

• a factor of 2 increase in the base case CDF

• a factor of 10 increase in the maximum attainable benefit 
-22-

DCD Tier 2, Table 1B-4, summarizes the results for these cases. With the exception of

the last sensitivity case, the calculated maximum attainable benefit was no more than $43,000. 

Even when the AP1000 CDF and LRF were increased by a factor of 10, the maximum

attainable benefit of eliminating all risk for the AP1000 increased to only about $200,000.

The applicant found that none of the 14 design alternatives and neither of the two

additional alternatives related to the PRA success criteria would be cost beneficial. Only one

alternative has an implementation cost close to $21,000, namely, SAMDA 3, self-actuating

CIVs, which has an estimated cost of $33,000. All of the remaining alternatives have estimated

implementation costs at least a factor of 20 greater than the maximum attainable benefit of

$21,000. On this basis, the applicant concluded that only SAMDA 3 warranted further

evaluation.

SAMDA 3 consists of improved containment isolation provisions on all normally open

containment penetrations. The design alternative would involve either adding a self-actuating

valve or enhancing the existing inside CIV to provide for self-actuation in the event that

containment conditions are indicative of a severe accident. Westinghouse noted that even if

this SAMDA completely eliminated all releases associated with containment isolation failures

(i.e., release category containment isolation (CI)) and reduced the CDF to zero, the benefit of

the SAMDA would be on the order of $1000. More realistically, the CDF would not be

impacted, and elimination of all containment isolation failures would only have a benefit on the

order of $100. Thus, even the lowest cost SAMDA would not be cost beneficial.

On the basis of the cost-benefit comparison, the applicant concluded that no additional

modifications to the AP1000 design were warranted.

4.6.2 NRC Evaluation

The applicant’s estimates of risk do not account for uncertainties either in the CDF or in

the offsite radiation exposures resulting from a core damage event. The uncertainties in both of
-23-

these key elements are fairly large because key safety features of the AP1000 design are

unique and their reliability has been evaluated through analysis and testing programs rather

than operating experience. In addition, the estimates of CDF and offsite exposures do not

account for the added risk from earthquakes.

As part of the AP600 review, the NRC did detailed analyses to assess design alternative

benefits, taking into account the uncertainties in estimated CDF, offsite releases of radioactive

materials from a severe accident, and the effects of external events. Given the similarities

between the AP1000 and AP600 design features and risk profiles and the sets of SAMDAs

relevant to each design, the NRC considers this prior evaluation for the AP600, summarized

below, to be applicable to the AP1000 as well.

The staff estimated the maximum benefits that could be achieved with the AP600 design

alternatives, assuming that a design alternative can either completely eliminate all core damage

events or completely eliminate offsite releases of radioactive materials in the event of a severe

accident. The estimates of benefits were calculated using the NRC-developed FORECAST

code (NUREG/CR-5595, Revision 1, “FORECAST: Regulatory Effects Cost Analysis Software

Manual, Version 4.1,” Science and Engineering Associates, Inc., July 1996). FORECAST

allows the use of uncertainty ranges for all key parameters and provides a means for combining

uncertainties in these parameters. For the purposes of estimating the maximum potential

benefit from the AP600 design alternatives, the staff assumed that external events and accident

sequences not yet accounted for in the PRA increased the reference CDF by two orders of

magnitude (i.e., a factor of 100).

The results of the analysis indicated that design alternatives which prevent accidents

(i.e., reduce the accident frequency to zero) are much more cost effective than design

alternatives which reduce or eliminate offsite releases, but which have no effect on accident

frequency. This is because of the fairly large benefits of averting onsite cleanup and
-24-

decontamination costs and avoiding replacement energy costs. Neither of these costs are

assumed to be impacted by design alternatives which do not reduce accident frequency. The

staff divided the design alternatives into two groups: those that impact the CDF and those that

impact containment performance, but not CDF. Benefits were estimated by taking the fractional

reduction in risk for each design alternative (compared to the AP600 baseline risk as defined by

the applicant) and applying that fraction to the mean benefits.

Design alternatives that were within a decade of meeting a benefit-cost criterion of

$5000/person-rem were subjected to further probabilistic and deterministic considerations. 

None of the design alternatives had a cost-benefit ratio of less than $5000/person-rem. The

only design alternatives which came within a decade of the $5000/person-rem criterion were

SAMDA 10, diverse IRWST injection valves, and SAMDA 3, self-actuating CIVs. The NRC

concludes, on the basis of further probabilistic and deterministic evaluations, that these design

alternatives are not cost beneficial and need not be further pursued.

Given the similarities between the AP1000 and the AP600 design features and risk

profiles and the sets of SAMDAs relevant to each design, the NRC considers the results of this

prior evaluation for the AP600 to be applicable to the AP1000 as well. Accordingly, the NRC

further evaluated these two SAMDAs for the AP1000, as discussed below.

4.7 Further Considerations

4.7.1 Self-Actuating Containment Isolation Valves

This design alternative would reduce the likelihood of containment isolation failure by

adding self-actuating valves or enhancing the existing CIVs for automatic closure when

containment conditions indicate a severe accident has occurred. Conceptually, the design

would either be an independent valve or an appendage to an existing fail-closed valve that

would respond to post-accident containment conditions. For example, a fusible link would melt
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in response to elevated ambient temperatures, venting the air operator of a fail-closed valve,

thus providing the self-actuating function. This design alternative is estimated to impact

releases from containment by less than 10 percent.

This improvement to the containment isolation capability would appear to be effective in

reducing offsite releases for accidents involving external and internal events. The addition of

this design alternative would impose minor operational disadvantages to the plant because the

operations and maintenance staff would require some additional training. These automatic

features would also require periodic testing to assure that they were functioning properly.

The most important question regarding this design alternative is whether it can be

implemented for a cost of only $33,000. The cost estimate appears not to include the first-time

engineering and qualification testing that would be required to demonstrate that the valve would

perform its intended function in a timely and reliable manner. The costs of periodic testing and

maintenance appear not to have been included. The NRC believes that the actual costs of this

design alternative would be substantially higher than the applicant’s estimate (by a factor of

10 or more) when all related costs are realistically considered. On the basis of the unfavorable

cost-benefit ratio and the expectation that actual costs would be even higher than the applicant

estimated, the NRC concludes that this design alternative is not cost beneficial and need not be

further evaluated.

4.7.2 Diverse IRWST Injection Valves

In the current AP1000 design, a squib valve in series with a CV isolates each of four

IRWST injection paths. This design alternative would reduce the likelihood of CCFs of IRWST

injection to the reactor by utilizing diverse valves in two of the four lines. The complete

elimination of the CCFs of IRWST injection squib valves would lead to a moderate (up to

10 percent) reduction of the at-power internal events CDF. In the absence of a comprehensive

external events PRA for the AP1000 plant, it is difficult to estimate the effectiveness of this
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design alternative in reducing the risk from external events such as seismic events. However, it

appears likely that failure to inject coolant to the reactor would remain a contributor to the CDF

from external events, in which case diversity in the IRWST injection valves should help to

reduce the risk from both external and internal events.

Alternate vendors are available for the CVs. However, it is questionable if CVs of

different vendors would be sufficiently varied to be considered diverse unless the type of CV 

was changed from the current swing-disk check valve type to another type. The swing-disk

type is preferred for this application and other types are considered less reliable.

Adding diversity to the injection line squib valves would require additional spares at the

plant and some additional training for plant operations and maintenance staff, but would not

appear to add significantly to the operational aspects of the AP1000. However, a greater issue

concerns the availability and costs of acquiring diverse valves from a second vendor. Squib

valves are specialized valve designs for which there are few vendors. The applicant claimed

that a vendor might not be willing to design, qualify, and build a reasonable squib valve design 

for this application, considering that the vendor would only supply two valves per plant. The

cost estimate for this design alternative assumes that a second squib valve vendor exists and

that the vendor only provides the two diverse IRWST squib valves per plant. The cost estimate

does not include the additional first-time engineering and qualification testing costs that will be

incurred by the second vendor. The applicant estimated that those costs could be more than

$1 million dollars. As a result, the applicant concluded that this design alternative would not be

practicable because of the uncertainty in the availability of a second squib valve design/vendor

and the uncertainty about the reliability of another type of CV. The NRC considers the rationale

set forth by the applicant regarding the potential reductions in reliability and high costs

associated with obtaining diverse valves to be reasonable. On the bases of these arguments,

the NRC concludes that this design alternative need not be further pursued.
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4.8 Conclusions

As discussed in Section 19.1 of the AP1000 FSER, Westinghouse used the PRA results

extensively to arrive at the final AP1000 design. As a result, the estimated CDF and risk

calculated for the AP1000 design are very low, both relative to existing operating plants and in

absolute terms. Moreover, the low CDF and risk for the AP1000 plant reflect Westinghouse’s

efforts to systematically minimize the effect of initiators/sequences that have been important

contributors to CDF in previous PWR PRAs. This minimization has been done largely through

the incorporation of a number of design improvements. Section 19.1 of the AP1000 FSER

discusses these improvements and the additional AP1000 design features which contribute to

low CDF and risk for the AP1000.

Because the AP1000 design already has numerous plant features designed to reduce

CDF and risk, the benefits and risk reduction potential of additional plant improvements is

significantly reduced. This reduction is true for both internally and externally initiated events. 

Moreover, with the features already incorporated in the AP1000 design, the ability to estimate

CDF and risk approaches the limitations of probabilistic techniques. Specifically, when CDFs

are estimated to be on the order of 1,000,000 years, it is possible that the areas of the PRA

where modeling is least complete, or supporting data are sparse or even nonexistent, may

actually be the more important contributors to risk. Areas not modeled or incompletely modeled

include human reliability, sabotage, rare initiating events, construction and design errors, and

systems interactions. Although improvements in the modeling of these areas may introduce

additional contributors to CDF and risk, the NRC does not expect that additional contributions

would change the conclusions in absolute terms.

The NRC concludes that none of the potential design modifications evaluated are

justified on the basis of cost-benefit considerations. The NRC further concludes that it is
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unlikely that any other design changes would be justified in the future on the basis of personrem exposure because the estimated CDFs are very low on an absolute scale. 

5.0 Alternative Use of Resources

No resources, such as land, water, or physical materials, will be affected by the

promulgation of this proposed rule. This proposed rule would codify the AP1000 design in the

Code of Federal Regulations but would not authorize the siting, construction, or operation of

any nuclear power plant.

6.0 States Consulted and Sources Used

The NRC has sent a copy of the proposed rule and draft EA to the State Liaison Officers

and requested their comments on the EA.

The Commission has determined under the NEPA of 1969, as amended, and the NRC’s

regulations in 10 CFR Part 51, Subpart A, that this rule is not a major Federal action

significantly affecting the quality of the human environment. Therefore, the NRC has

determined that preparation of an environmental impact statement for this rulemaking is not

required. The basis for this determination, as documented in this EA, is that the amendment to

10 CFR Part 52 would not authorize the siting, construction, or operation of a facility referencing

the AP1000 design; it would only codify the AP1000 design in a rule. Therefore, the NRC staff

did not issue this EA for comment by Federal, State, and local agencies. The NRC’s finding of

no significant environmental impact was published in the Federal Register on xxxxxxxx, with the

proposed design certification rule and draft EA for the AP1000 design. The NRC will evaluate

the EIs and issue an EIS as appropriate in accordance with NEPA as part of the application(s)

for the siting, construction, or operation of a facility.
-29-

FINDING OF NO SIGNIFICANT IMPACT:

On the basis of the environmental assessment, the NRC concludes that the proposed

action will not have a significant effect on the quality of the human environment. Accordingly,

the NRC has decided not to prepare an environmental impact statement for the proposed

action.

For further details with respect to the proposed action, see the proposed design

certification rule. Documents may be examined, and/or copied for a fee, at the NRC’s Public

Document Room (PDR), located at One White Flint North, 11555 Rockville Pike (first floor),

Rockville, Maryland. Publicly available records will be accessible electronically from the

Agencywide Documents Access and Management System (ADAMS) Public Electronic Reading

Room on the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. Persons who do not

have access to ADAMS or who encounter problems in accessing the documents in ADAMS

should contact the NRC PDR reference staff at 1-800-397-4209 or 301-415-4737 or send an 

e-mail to pdr@nrc.gov.

Note: Public access to documents, including access via ADAMS and the PDR, has been

temporarily suspended so that security reviews of publicly available documents may be

performed and potentially sensitive information removed. However, access to the

documents identified in this rule continues to be available through the rulemaking web

site at http://ruleforum.llnl.gov, which was not affected by the ADAMS shutdown. Please 
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check with the listed NRC contact concerning any issues related to document

availability.

Dated at Rockville, Maryland, this day of

FOR THE NUCLEAR REGULATORY COMMISSION

William D. Beckner, Program Director

New, Research and Test Reactors Program

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation
STATE LIAISON OFFICERS

SUBJECT: AP1000 DESIGN CERTIFICATION

The United States Nuclear Regulatory Commission (NRC) has sent to the Office of the

Federal Register for publication the enclosed proposed amendment to the Commission’s rules

in Title 10 of the Code of Federal Regulations (10 CFR) Part 52. This amendment resulted

from Westinghouse Electric Company’s (Westinghouse’s) request for certification of its AP1000

standard plant design on March 28, 2002. The NRC staff issued a final design approval to

Westinghouse on September 13, 2004, that signified completion of the technical review phase

and readiness for the rulemaking phase of the AP1000 application. 

This proposed rule would amend 10 CFR Part 52 to certify the AP1000 standard plant design.

This proposed rulemaking action is necessary so that applicants or licensees intending to

construct and operate an AP1000 design may do so by referencing the AP1000 design

certification rule (DCR). 

Also enclosed is an environmental assessment (EA) that has been prepared in support of the

proposed rule. The conclusion of the EA is the Commission’s finding that no significant

environmental impact will result from the proposed rule. The EA and the Federal Register

Notice (FRN) are provided for your review and comment. If you have any comments on the rule

and its environmental impact, please send them by [insert the date, using the same number of

days allowed for public comments in the FRN]. Comments received after this date will be

considered if it is practical to do so, but the Commission is able to assure consideration only for

comments received on or before this date.

You can use the following methods to transmit your comments: (1) mail your written comments

to the Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,

Attention: Rulemakings and Adjudications Staff; (2) fax your comments to (301) 415-1101;

and (3) provide comments via the NRC’s interactive rulemaking Website at

http://ruleforum.llnl.gov. This site provides the capability to upload comments as files (any

format) if your Web browser supports that function. For information about the interactive

rulemaking Website, contact Ms. Carol Gallagher, (301) 415-5905 (e-mail: CAG@nrc.gov).

For additional information on this action, please contact Lauren M. Quinones-Navarro,

Mailstop O-4D9A, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, telephone (301) 415-2007, e-mail LNQ@nrc.gov.

Sincerely,

Paul H. Lohaus, Director

Office of State and Tribal Programs

Enclosure: As stated