NUREG-1789

10 CFR Part 52 Construction

Inspection Program Framework

Document

U.S. Nuclear Regulatory Commission POR

Office of Nuclear Reactor Regulation

Washington, DC 20555-0001 ; Rev
AVAILABILITY OF REFERENCE MATERIALS

IN NRC PUBLICATIONS

NRC Reference Material

As of November 1999, you may electronically access

NUREG-series publications and other NRC records at

NRC's Public Electronic Reading Room at

htt: '/www.nrc.gov/reading-rm.html.

Publicly released records include, to name a few,

NUREG-series publications; Federal Register notices;

applicant, licensee, and vendor documents and

correspondence; NRC correspondence and internal

memoranda; bulletins and information notices;

inspection and investigative reports; licensee event

reports; and Commission papers and their attachments.

NRC publications in the NUREG series, NRC

regulations, and Title 10, Energy, in the Code of

Federal Regulations may also be purchased from one

of these two sources.

1. The Superintendent of Documents

U.S. Government Printing Office

Mail Stop SSOP

Washington, DC 20402-0001

Internet: bookstore.gpo.gov

Telephone: 202-512-1800

Fax: 202-512-2250

2. The National Technical Information Service

Springfield, VA 22161-0002

www.ntis.gov

1-800-553--6847 or, locally, 703-605-6000

A single copy of each NRC draft report for comment is

available free, to the extent of supply, upon written

request as follows:

Address: Office of the Chief Information Officer,

Reproduction and Distribution

Services Section

U.S. Nuclear Regulatory Commission

Washington, DC 20555-0001

E-mail: DlSTRIBUTION~nrc.gov

Facsimile: 301-415-2289

Some publications in the NUREG series that are

posted at NRC's Web site address

httD://www. nrc Qov/reading-rm/doc-collections/nureQs

are updated periodically and may differ from the last

printed version. Although references to material found

on a Web site bear the date the material was accessed,

the material available on the date cited may

subsequently be removed from the site.

l

Non-NRC Reference Material

Documents available from public and special technical

libraries include all open literature items, such as

books, journal articles, and transactions, Federal

Register notices, Federal and State legislation, and

congressional reports. Such documents as theses,

dissertations, foreign reports and translations, and

non-NRC conference proceedings may be purchased

from their sponsoring organization.

Copies of industry codes and standards used in a

substantive manner in the NRC regulatory process are

maintained atThe NRC Technical Library

Two White Flint North

11545 Rockville Pike

Rockville, MD 20852-2738

These standards are available in the library for

reference use by the public. Codes and standards are

usually copyrighted and may be purchased from the

originating organization or, if they are American

National Standards, fromAmerican National Standards Institute

11 West 42nd Street

New York, NY 10036-8002

www.ansi.org

212-642-4900

Legally binding regulatory requirements are stated

only in laws; NRC regulations; licenses, including

technical specifications; or orders, not in

NUREG-series publications. The views expressed

in contractor-prepared publications in this series are

not necessarily those of the NRC.

The NUREG series comprises (1) technical and

administrative reports and books prepared by the

staff (NUREG-XXXX) or agency contractors

(NUREG/CR-XXXX), (2) proceedings of

conferences (NUREG/CP-XX)XX), (3) reports

resulting from international agreements

(NUREG/IA-XXXX), (4) brochures

(NUREG/BR-XXXX), and (5) compilations of legal

decisions and orders of the Commission and Atomic

and Safety Licensing Boards and of Directors'

decisions under Section 2.206 of NRC's regulations

(NUREG-0750).
ABSTRACT

The information contained in the Construction Inspection Program Framework Document, NUREG1789, details the overall philosophy and approach that the NRC will use to inspect new nuclear

power reactors being licensed and built under 10 CFR Part 52. The information detailed in this

NUREG about the construction of new reactors will guide the development of inspection manual

chapters and inspection procedures that will be used to implement the construction inspection

program.

iii
CONTENTS

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

EXECUTIVE SUMMARY .................... ix

ACKNOWLEDGMENTS . ....................... . xi

ABBREVIATIONS .......................... xiii

1. INTRODUCTION ................... 1

1.1 History ........................ 1

1.2 10 CFR Part 52 Process ................- 2

1.3 Expected Licensing and Construction Environment ................ ........... 3

2. FRAMEWORK OVERVIEW ........................................... 6

3. FRAMEWORK IMPLEMENTATION .8

3.1 IMC-2501, 'Early Site Permit".. 8

3.1.1 Introduction. 8

3.1.2 Audits/inspections .8

3.1.2.1 Pre-Application Audits. 8

3.1.2.2 Post Application Inspections. 9

3.1.3 Enforcement. 9

3.1.4 QualityAssurance. 9

3.2 IMC-2502, 'Construction Inspection Program: Pre-Combined License (Pre-COL)

Phase . . 10

3.2.1 Introduction ......... 10

3.2.2 Inspections ......... 10

3.2.2.1 Quality Assurance Inspections During the Preparation of

the Application ....... ; - 11

3.2.2.2 Engineering Design Verifications and First-of-a-Kind Engineering

Inspections ....... 11

3.2.2.3 Operational Program Reviews and Inspections ........ .......... 12

3.2.2.4 Inspections of Other Activities Completed During the Pre-COL Phase 12

3.2.3 Construction Inspection Program Information Management System (CIPIMS) 13

3.2.4 Enforcement . ............................................. 13

3.3 IMC-2503, 'Construction Inspection Program: Inspections, Tests, Analyses, and

Acceptance Criteria ................. ................... 13

3.3.1 Introduction ...................... 13

3.3.2 ITAAC Inspection Overview ...................... 14

3.3.3 ITAAC Inspection Philosophy . ; ....................... 16

v
3.3.4 ITAAC Inspection Process ......................................... 16

3.3.4.1 Inspection Results .................. ....................... 16

3.3.4.2 Review of Inspection Results ............... .................. 17

(1) Positive SAYGO Determinations ........................... 17

(2) Negative SAYGO Determinations .......................... 17

3.3.4.3 ITAAC Determinations .. 18

(1) Determining ITAAC Acceptability .......................... 19

(2) Invalidation of Previously Accepted ITAAC Determinations ...... 19

3.3.4.4 Commission 10 CFR 52.103(9) ITAAC Finding ....... ........... 19

3.3.4.5 Public Notifications .................. ....................... 20

3.3.4.6 Enforcement ..................... ......................... 20

3.3.5 Inspecting Module Construction Activities ............................. 20

3.4 IMC-2504, "Construction Inspection Program: Non-ITAAC Inspections .... ....... 22

3.4.1 Introduction .................................................... 22

3.4.2 Before 10 CFR 52.103(g) Finding ................................... 23

3.4.2.1 Inspection After an ITAAC is Met and Prior to Fuel Loading .... .... 23

3.4.2.2 Operational Readiness Assessments .......................... 24

3.4.3 Post Fuel Load Prior to Power Operations ............................ 25

3.5 Construction Inspection Program Information Management System (CIPIMS) ..... 25

3.6 NRC Organization ..................................................... 27

3.6.1 Introduction .................................................... 27

3.6.2 Implementation .................................................. 27

3.6.2.1 ESP .................................................... 27

3.6.2.2 COL and Beyond .......................................... 27

3.6.3 Items To Be Considered During the Development of the

Inspection Organization ........................................... 27

3.6.3.1 Regional Involvement ...................................... 28

3.6.3.2 Inspection Scheduling and Data Management Activities .... ....... 28

3.6.3.3 Inspection Project Management .............................. 29

3.6.3.4 ITAAC Comparison to Inspection Procedures ..... .............. 30

3.6.6.5 Inspection Findings and Enforcement ......................... 30

4. PUBLIC INVOLVEMENT ................................................... 32

Appendix A Glossary .A-1

Appendix B Information Considered in Updating the Construction Inspection Program ... B-1

Appendix C Inspection Sampling .C-1

Appendix D Examples of Information That Would Invalidate a Previous

ITAAC Determination .D-1

vi
Appendix E

Appendix F

Appendix G

Appendix H

Appendix I

ABWR Construction Example ...................... E-1

Generic Combined License From SECY-00-0092 .F-1

An example of an NRC Organization for Implementing a Construction

Inspection Program .G-1

Comparison of AP600 and ABWR ITAAC to Inspection Procedures .H-1

Comment Resolution Summary for the Draft Construction Inspection Program

Framework Document 1.-

vii
EXECUTIVE SUMMARY

This framework document will be used as the guiding document for the creation of construction

inspection manual chapters and inspection procedures to support the 10 CFR Part 52 licensing

process. The staff initially published this document in May 2003 to solicit stakeholder comments.

After receiving and considering stakeholder comments, the staff has revised the document and is

issuing NUREG-1789 to describe the final construction inspection framework.

In 1991, the Office of Nuclear Reactor Regulation (NRR) started revising the Construction

Inspection Program (CIP) governed by Inspection Manual Chapter (IMC) 2512, "Light Water

Reactor Inspection Program - Construction Phase." This project had two purposes: to address

NRC construction inspection programmatic weaknesses that had been identified during the

licensing of several plants, and to develop an inspection program for evolutionary and advanced

reactors. The project was stopped in the mid 1990s because no new nuclear power plants were

being constructed. A report was assembled that provided an approach reactivating a future

construction inspection program. The "Draft Report on the Revised Construction Inspection

Program," was issued in October of 1996.

In SECY-01-0188, "Future Licensing and Readiness Assessment (FLIRA)," the NRR' staff

recognized the need to resume revising the CIP and, in 2001 a CIP team consisting of NRR and

regional inspectors was established to do so. A steering committee consisting of NRR and regional

managers was also formed.

This CIP framework document updates the work that was published in 1996. It incorporates the

applicable requirements of 10 CFR Part 52 such as inspections, tests, analyses, and acceptance

criteria (ITAAC), and provides allowances for the rapid construction schedules made possible by

the parallel and modular construction techniques used in today's construction environment.

Whether or not a combined license (COL) applicant references a certified design, the COL

application will contain ITMC. Should the Commission grant the COL, the staffs construction

inspection activities will need to be organized to document the staffs determinations on the

licensee's completion of ITAAC.

It is expected, however, that inspections for plants licensed in accordance with 10 CFR Part 52 will

be conducted in the same manner as during the construction of earlier reactors, using the same

procedures revised, updated, and supplemented by new administrative inspection processes. The

Part 52 CIP, therefore, is essentially an updated revision of the older NRC construction inspection

program previously used to inspect all light water reactors licensed under 10 CFR Part 50. An

electronic information tracking and scheduling system (the CIP Information Management System)

is being developed to coordinate inspections with licensee construction schedules and to facilitate

access, retrieval, and tracking of inspection findings, reviews, and determinations about the

licensee's ITAAC. Appendix A contains a glossary for terms that are used throughout this report.

The program has four phases. The first phase supports a licensing decision for an early site permit

(ESP), the second phase supports issuance of a combined license (COL), and the third and fourth

phases support construction activities and the preparations for operations.

ix
Inspections will initially be performed to confirm the accuracy of data submitted to the NRC in

support of safety evaluations and the Atomic Safety and Licensing Board (ASLB) hearings for an

ESP and COL. Because work, such as data collection and procurement, may take place before the

staff receives an ESP or COL application, the inspection activities could begin prior to an

application. During plant construction, the focus will shift to verifying satisfactory completion of

ITAAC, as specified in the final safety analysis report (FSAR), and also to inspecting programs for

operational readiness and transition to power operations.

With ITAAC structured as they are, the staff will need to make determinations regarding the

completion of individual ITAAC, as the licensee indicates completion of them. Therefore, a phased

verification program was developed to assess completion of activities. A sign-as-you-go (SAYGO)

methodology will be used, beginning at the early stages of construction; the staff will publish their

determinations on individual ITAAC as they are completed; and finally, after all ITAAC have been

completed, the Regional Administrator, through the Director of NRR, will make a recommendation

to the Commission about whether or not the ITAAC have been met.

Inspectors will verify the completion of ITAAC and document their determinations in inspection

reports which will be published on an NRC Web site. At appropriate intervals during construction,

ITAAC determinations will be published in the Federal Register. It is expected that most negative

inspection findings will be resolved primarily by the licensee's corrective action program, but more

significant inspection findings may require NRC management involvement. All inspection findings,

assessments, ITAAC determinations, and open items will be tracked by the CIP Information

Management System.

Since the NRC has limited resources and uses a sampling inspection methodology, reduction in

inspection effort may occur when reviews have identified effective program implementation that

provides high confidence in the licensee's quality control process. Such reviews will determine

whether construction process controls associated with a particular activity are satisfactory. Initially,

the activity is heavily inspected. If the process controls are found acceptable, the resources are

reduced and that activity is inspected less frequently. Examples of these types of activities are

welding, cable pulling, installing pipe supports, and installing electrical penetrations.

The NRC will conduct inspections of operational programs. The scope of the inspections will be

similar to the scope of the previous construction program. Most of these NRC inspection findings

will be resolved by the licensee's corrective action program, or by traditional enforcement measures.

All inspection findings are entered into the NRC CIP Information Management System for easy

access, tracking, sorting, and retrieval.

x
ACKNOWLEDGMENTS

This framework document was developed by a team of individuals from the Nuclear Regulatory

Commission's headquarters and regional offices. In addition, a steering group consisting of

headquarters managers and a regional manager provided comments and guidance for this

document. The construction inspection program steering group is composed of the following

managers:

Stuart A. Richards, Chief, Inspection Program Branch, NRR

James E. Lyons, Program Director, New, Research and Test Reactors, NRR

Charles A. Casto, Director, Division of Reactor Safety, Region 11

The construction inspection program team that finalized this document is composed of the following

members:

Mary Ann Ashley, Inspection Program Branch, NRR

Jerome Blake, Region 11

Antone Cerne, Region I

Thomas Foley, Inspection Program Branch, NRR

Ronald Gardner, Region Ill

Caudle Julian, Region 11

Edmund Kleeh, Inspection Program Branch, NRR

Charles Paulk, Region IV

Joseph Sebrosky, New Reactor Licensing Section, NRR

The team would also like to acknowledge the contributions of Michele Evans, Jim Isom, and Carl

Konzman during the development of the draft framework document.

xi
ABBREVIATIONS

ABWR Advanced Boiling Water Reactor

ADAMS Agencywide Documents Access and Management System

ASLB Atomic Safety and Licensing Board

CIP Construction Inspection Program

CIPIMS Construction Inspection Program Information Management System

COL Combined License

DAC Design Acceptance Criteria

ESP Early Site Permit

FLIRA Future Licensing Inspection and Readiness Assessment

IMC Inspection Manual.Chapter

ITAAC Inspections, Tests, Analyses, and Acceptance Criteria

LWA Limited Work Authorization

NRC U.S. Nuclear Regulatory Commission

NRR Office of Nuclear Reactor Regulation

ORAT Operations Readiness Assessment Team

PM Project Manager

PRA Probabilistic Risk Assessment

QA Quality Assurance

QC Quality Control

ROP Reactor Oversight Program

SAYGO Sign As You Go

SER Safety Evaluation Report

SSC System, Structure, or Component

WPC Work Planning Center

xiii
1. INTRODUCTION

1.1 History

In 1991, an NRR working group was established to revise the existing IMC-2512, Light Water

Reactor Inspection Program Construction Phase, to incorporate lessons learned from previous

construction experience and to develop a program to inspect evolutionary and advanced reactors

that might be licensed under 10 CFR Part 52.

At the start of program development in 1991, the working group collated the construction inspection

experience within the- NRC. The working group revised the inspection programs, policies, and

structure and issued a draft report on the revised construction inspection program (CIP). At the

same time, the working group developed a computer-based inspection scheduling system to assist

the NRC staff in implementing the CIP.' The final computer system was based on a system

completed for Bellefonte nuclear plant but was also intended for deployment at other nuclear power

plants under construction. Because of the lack of new reactor construction activities, the inspection

scheduling system was never field-tested. Subsequently, the staff refined the system and called

it the CIP Information Management System (CIPIMS).

The CIPIMS provided enhanced guidance and capabilities for gathering, recording, and reporting

construction inspection information. The enhancements involved the use of a systems-based

inspection planning methodology, the computerization of the inspection program, and a continuous

onsite inspection presence throughout plant construction.

The development of the computer program continued until the mid 1990s, when the project was

suspended because of NRC staff resource constraints and a lack of nuclear power plant

construction. CIPIMS is described in the "Draft Report on the Revised Construction Inspection

Program," dated October 1996. In addition to describing CIPIMS, the report presented a framework

for the reactivation of a future construction inspection program.

In SECY-01-0188, "Future Licensing and Inspection Readiness Assessment," dated October 12,

2001, the staff recognized the need to restart this effort. The Office of Nuclear Reactor Regulation

(NRR) established a CIP team and steering group and began routine public meetings to discuss

construction inspection related issues. The CIP team used the 1996 draft report on the revised CIP

as a framework to reactivate the dormant CIP.

The team recognized that several assumptions about the construction inspection program have

changed since 1996. For example, the 1996 draft report on the revised CIP was written so it could

be used either for a 10 CFR Part 50 licensing process or for the newer 10 CFR Part 52 licensing

process. While the 10 CFR Part 50 process can still be used to license and construct a nuclear

power plant, the industry has indicated that it does not intend to use this process. Additionally, the

main focus of the draft report was on the inspection of actual construction activities because that

is where most of the work in revising the construction inspection program needed to be done. Early

site permit inspection guidance was only given a slight mention in the draft report of 1996. The

team recognized that detailed early site permit (ESP) inspectioin'guidance needed to be developed

rapidly in order to support the ESP application schedules proposed by industry.

1
Based on the changes to the CIP which have occurred since 1996, the team decided to update the

CIP and to issue new manual chapters for inspections under the 10 CFR Part 52 process. These

new inspection manual chapters (IMCs) are based on previous guidance in the IMCs. used to

assess construction activities for nuclear power plants constructed in accordance with 10 CFR Part

50. The new IMCs also consider and incorporate many of the lessons learned that are discussed

in Appendix B of this document.

1.2 10 CFR Part 52 Process

Future U.S. nuclear power plants will be licensed under either 10 CFR Part 50 or 10 CFR Part 52.

The new inspection manual chapters have been structured to accommodate the .10 CFR Part 52

licensing process. A brief description of the 10 CFR Part 52 licensing process is provided below

in order to place the various construction inspection manual chapters in context.

In 1989, the NRC established new alternatives for nuclear plant licensing under 10 CFR Part 52.

Part 52 describes a combined licensing process, an ESP process, and a standard plant design

certification process. This approach allows early resolution of safety and environmental issues.

The issues resolved by the design certification rulemaking process and during the ESP hearing

process are not reconsidered during! the COL review except under narrow, clearly defined

circumstances. Figure 1.1, TCombined M Licenses, Early- Site Permits and Standard Design

Certifications," below shows the relationship among the three processes.

e_ S a- .*...S...... "

.: `..

Veiffcotron of inspection.

TEatm. Anndv.as and

Acceptance Crtera

Standard Design ~

Cer.fl_ lon =4Reactrpemt

I~ - -

: Imcfo- - 9\ 4 4 

Combined License

Review and Hearing

of OQuIVo;nU proe

Figure 1.1. Combined Licenses, Early Site Permits and Standard Design Certifications

2
The construction inspection manual chapters provide guidance for the activities in the 10 CFR Part

52 process. The following paragraphs briefly describe ESPi, standard design certifications, and

COLs.

* Early Site Permits

Under the NRC's regulations in 10 CFR Part 52, the agency can issue an ESP for approval of one

or more sites separate from an application for a construction permit or COL. Such permits are good

for ten to twenty years and can be renewed for an additional ten to twenty years. They address site

safety issues, environmental protection issues, and plans for coping with emergencies, and are

independent of the staffs review of a specific nuclear plant design. Because this is a new process,

IMC-2501 and inspection procedures were developed and issued to provide guidance for

inspections to be performed to support the issuance of an ESP.

* -Standard Design Certification

The NRC can certify a reactor design forfifteen years through the rulemaking process, independent

of a specific site. As set forth in 10 CFR 52.47, an application for a standard design certification

must contain information describing the design and proposed inspections, tests, analyses, and

acceptance criteria (ITAAC) that are necessary and sufficient to provide reasonable assurance that,

if the ITAAC are performed and the acceptance criteria met, a plant which references the design

is built and will operate in accordance with the design certification.

* Combined License (COL)

A COL authorizes construction and operation of a nuclear power plant with conditions. The

application for a COL must contain essentially the same information required in an application for

an operating license submitted under 10 CFR Part 50, including financial and antitrust information

and an assessment of the need for power. The application must also describe the ITAAC that are

necessary to ensure that the plant has been properly constructed and will operate safely.

An application for a COL may reference a standard design certification, an ESP, both, or neither.

If the application references a standard design certification, the applicant must perform the ITAAC

for the certified design and the site-specific design features. If the application does not reference

a standard design certification, the applicant must provide complete design information, including

certain information that the applicant would otherwise have submitted for a standard design

certification. Similarly, if the application does not reference an ESP, the applicant must provide

detailed siting information that would otherwise have been provided during the ESP process.

Should the NRC issue a COL, the NRC then verifies that the licensee has completed the required

ITAAC and that the ITAAC acceptance criteria have been met before the plant can operate.

1.3 Expected Licensinq and Construction Environment

New certified designs with accelerated construction schedules are being marketed to improve the

overall cost effectiveness of nuclear power generation. The accelerated construction schedules are

3
based, in part, on the modular design of these reactors. For the new generation of light water

reactors, and for the gas-cooled reactor, the staff understands that applicants plan to have many

of the systems/subsystems fabricated at remote facilities (e.g., U.S. or foreign-based shipyards),

then to ship these systems/subsystems to the facility for construction of the unit(s). Figure 1.2,

"Modular Construction Diagram," provides a representation of this modular construction technique.

The staff has had discussions with several different vendors and all plan to use these techniques

for plants constructed in the U.S. The nuclear power industry in China and Japan is also currently

employing modular construction techniques.,

Under 10 CFR Part 52, fabrication activities can begin even before an applicant announces its intent

to submit an application for a COL. Therefore, fabrication activities could begin off-site prior to

Commission approval of a COL application, but will most likely not begin before an applicant

submits an application for a COL.,

The expected rapid pace of future nuclear power plant construction will call for the NRC to schedule

inspection activities in a way that will ensure that construction inspection does not become a critical

path-activity. While not required by regulations; an applicant-should notify, the NRC before

fabrication activities begin, to allow the staff sufficient time to- plan and implement inspection

activities. To assist in more effective inspection scheduling, the applicant's construction plan should

- be incorporated, if possible, into the construction inspection schedule. -Close coordination between

inspection and construction schedules will be needed.

___ ___ ___ ___ __ n .

RailiTruck Shipment of Modules

Factory Production - of Modules

nt Order

She Survey

and Preparation

Onsite Module

Assembly

M .. X .

Site Construction

Time - !

Figure 1.2 - Modular Construction Diagram; -

4
Depending on the extent of modular construction employed, the inspection staff may need to verify

ITAAC at remote facilities during the pre-COL phase. In general, however, critical attributes of

systems, structures, and components (SSCs) should be inspected on site to the maximum extent

possible. Scheduling inspections at fabrication facilities may be difficult but is important since the

fabrication of modules and major plant components could begin many months before the COL is

issued and before the first structural concrete is placed.

Evaluations of modules intended to be installed into a plant will be conducted to identify potential

modes of degradation during transit.

5
2. FRAMEWORK OVERVIEW

The Part 52 CIP is essentially an updated revision of the older NRC construction inspection program

previously used to inspect all light water reactors built in the United'States under 10 CFR Part 50.

The old program consisted of five inspection manual chapters; The Part 52 CIP modifies each of

these manual chapters to incorporate the applicable requirements of 10 CFR Part 52. The same

types of inspections are expected to be conducted for a plant licensed in accordance with Part 52

as those performed during the construction of earlier reactors.- These inspections will use the same

procedures, as revised, reorganized, updated and supplemented by new administrative inspection

processes and the CIPIMS, an electronic information tracking and scheduling system. The CIPIMS

was developed to coordinate the inspection and licensee construction schedules and to track

inspection findings and ITAAC completion status.

Inspections will focus on two areas: (1) verifying satisfactory completion of ITAAC as specified in

the final safety analysis report and (2) compliance with regulations (e.g., Part 21; 10 CFR 50,

Appendix B; 10 CFR 50.55(e)), which may not be directly related to an ITAAC. If the inspections

identify deficiencies such that an ITAAC was no or will not be met, or that the licensee is not in

compliance with regulations the staff will document and docket the information and make it publicly

available. The licensee will be expected to take appropriate corrective actions to address

deficiencies. The staff may take enforcement action, as appropriate, for instances where

regulations have not been met. If the deficiency is not corrected, the Commission may elect not to

authorize fuel load. The IMC guidance, for the most part, is written assuming that the licensee will

be able to correct such deficiencies.

The staff will verify the completion of certain ITAAC by simply comparing system performance

measurements and observations against established criteria. ITAACs of this type will normally be

accomplished within a well-defined period during construction and their completion will be easily

documented. The licensee will complete the inspection, test, and analysis for other ITAAC, such

as welding, over a longer period during construction, and the NRC will perform many inspections

to verify their various attributes. When the final construction activity for this type of ITAAC is

completed, results of the NRC inspections will contribute to a staff determination of the successful

completion of the activity and will ultimately support a Commission finding that all of the ITAAC have

been met. In order to allow timely verification of ITMC that will be done over a long period of time,

a "Sign As You Go' (SAYGO) method will be used. This method will require NRC inspectors to sign

off completed ITAAC, or portions of complex ITAAC, early in the process as they are successfully

demonstrated to the inspector, hence, sign-as-you-go. In accordance with 10 CFR 52.99, at

appropriate intervals during construction, the NRC will publish notices of the successful completion

of ITAAC in the Federal Register. The staffs inspection findings and assessments with respect to

ITAAC will be published in inspection reports. This method will provide an on-going record of the

acceptability of the work related to the ITAAC.

The new CIP inspection manual chapters can be broken into two categories: (1) Chapters that are

done in order to support a licensing decision (i.e., IMC-250 1 for an ESP, and IMC-2502 for issuance

of a COL), and (2) chapters that are done to verify aspects of construction activities and to provide

for the transition to the operations phase (i.e., IMC-2503 and IMC-2504).

6
The following inspection manual chapters will provide guidance for various inspection activities

during the reactor construction period:

(1) IMC-2501, uEarly Site Permit," was issued on October.8,2002. It is implemented when the

NRC is formally notified that an applicant is preparing an application for an ESP. It provides

guidance for staff inspection activities from that time, through the receipt of an ESP

application, to the subsequent safety review in support of the mandatory hearing that will

lead up to the Commission's decision to approve or disapprove the application for an ESP.

(2) IMC-2502, 'Pre-Combined License (Pre-COL) Phase," will provide guidance for inspection

activities from the time the NRC is notified of a person's intent to apply for a COL, through

the receipt of an application for a COL, to the mandatory hearing that leads to the

Commission's decision to approve or disapprove an application for a COL. This IMC and

the associated inspection procedures (IPs) will be used to facilitate the inspection activities

necessary to support the safety review leading up to the public hearing. The staff will also

use this guidance in overseeing any construction activities permitted under § 52.91.

(3) IMC-2503, "Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)," will provide

guidance for inspection activities to support the staffs review of the licensee's claim that

ITAACs have been met. The results of the staffs review will support the Commission's

decision on whether to allow fuel loading for a facility that has an approved COL.

(4) IMC-2504, 'Non-ITAAC Inspections," will provide guidance for inspection activities for the

requirements of 10 CFR Part 52 and the pre-operational testing phase. Prior to fuel load,

IMC-2504 will provide guidance for inspections other than ITAAC. After fuel load, the

guidance in this IMC will be used for inspections during initial fuel load, startup and power

ascension testing and will be used to guide the transition to IMC-2515.

Inspections associated with IMC-2503 and IMC-2504 will be conducted in parallel and could start

at placement of contracts for major component and module manufacturing. However, inspections

associated with IMC-2503 end at fuel load while those related to IMC-2504 will end when IMC-2515

is fully implemented.

7
3. FRAMEWORK IMPLEMENTATION

3.1 IMC-2501. "Early Site Permit"

3.1.1 Introduction

The requirements and procedures for approval of a site or sites for one or more nuclear power

facilities are defined in 10 CFR Part 52, Subpart A, "Early Site Permit" (ESP). Inspection Manual

Chapter (IMC) 2501 establishes guidance for NRC inspection activities directed towards both preESP application audit activities and post-application inspection activities.

The principal regulatory objective of the ESP phase is to verify that the ESP application meets the

requirements specified in 10 CFR Part 52. The application for an ESP must address three areas.

First, it must provide a description and safety assessment of the suitability of the site on which the

facility is to be located, including the seismic, meteorological, hydrological, and geologic

characteristics of the site. Second, it must provide a complete environmental report. Third, it must

identify physical characteristics unique to the proposed site that could pose a significant impediment

to the development of emergency plans. In addition, the applicant, at its option, may propose major

features of the emergency plan or complete an integrated emergency plan for NRC review in

consultation with the Federal Emergency Management Agency.

The NRC staff should also explain to the public the contents of an ESP application, the NRC

licensing and enforcement process,- and the opportunities for public participation. Finally the staff

should ascertain whether the elements and standards appropriate to assure quality are being

applied to the applicant's ongoing project activities.

Inspections and audits are conducted to verify the quality and accuracy of data collected and the

analysis and the evaluation of information used in support of the ESP application.

3.1.2 Audits/lnspections

The ESP phase for a plant is implemented when the NRC receives written notification of an

applicant's intention to apply for an ESP under 10 CFR Part 52. The inspection program is

applicable to the applicant and the applicant's contractors and to all activities related to NRC

regulations.

3.1.2.1 Pre-Application Audits

Before receiving the application, and as soon as possible after being notified of the applicant's

intention to submit an ESP, the staff holds meetings with the applicant to establish the primary

contacts forthe various technical disciplines, review the applicant's schedule for data collection, and

arrange to observe the applicant's implementation of its data collection program. The meetings are

8
also used to arrange a preliminary walk-down of the prospective site and to review the applicant's

controls for assuring quality in the application. The NRC coordinates schedules with the applicant

and gathers information in preparation for public meetings, schedules public meetings to introduce

the local community to the NRC licensing process, and arranges meetings with State and local

officials.

During the pre-application phase, the NRC conducts audits of the applicant's pre-application

activities to check for problems that could lead to an application being rejected. The audits gather

information primarily regarding the quality of site suitability data and environmental data collected

and the quality of analytical methodologies used in support of the application.

3.1.2.2 Post Application Inspections

During the post-application period, inspections are conducted to support the staff's safety evaluation

report testimony for the Atomic Safety and Licensing Board (ASLB) hearing required by 10 CFR

52.21. Based on the information provided by the applicant and the results of the inspections, safety

evaluation reports (SERs) are issued and the ASLB hearing is conducted prior to making a

determination on whether to grant the ESP.

Inspections are accomplished by the regional office having geographical jurisdiction over the

proposed site, with technical support from NRR. Inspections are led by the responsible region after

coordinating the effort with the responsible NRR project manager (PM). Technical support is

provided by various divisions within NRR as requested by the PM. The technical staff evaluates

the applicant's methodologies for data collection. Inspections are consolidated, when practicable,

to minimize impact on the applicant. Shortly after the conclusion of each inspection, the NRR

technical staff forwards its findings to the inspection team leader for integration into an inspection

report.

The inspection procedure guidance in Enclosure 1 to MC-2501 provides the inspector with the

applicable inspection procedures for use during inspections, audits, or site visits.

3.1.3 Enforcement

Enforcement actions associated with an ESP application are not anticipated in the pre-application

phase. However, as stated in Section 52.21, an ESP is a partial construction permit and is therefore

subject to all procedural requirements in 10 CFR Part 2 applicable to construction permits. The

information submitted with the application is subject to NRC regulations, including enforcement

actions for incomplete or inaccurate information.

3.1.4 Quality Assurance

During ESP activities the applicant should implement QA measures that are equivalent in substance

to the requirements of 10 CFR Part 50, Appendix B. This is necessary because the Commission,

in proceedingspon construction permits, operating licenses, or COLs, treats as resolved those

matters resolved during the ESP proceedings, as required by 10 CFR 52.39. Because of this

9
finality, conclusions derived during the ESP phase will be relied upon for use in subsequent design,

construction, fabrication, and operation of a reactor that might be constructed on a site for which

an ESP has been issued. Therefore, the quality measures implemented for activities important to

safety should be equivalent during the ESP and COL phases.

ESP activities associated with site safety assessment should be controlled by QA measures

equivalent in substance to the controls described in Appendix B to 10 CFR Part 50. The site safety

assessment establishes information, such as analyses and data, that is material to the reliable

performance of SSCs important to safety and will be used in the design, construction, and operation

of reactor systems that might be constructed on the proposed site. The QA measures provide

adequate confidence that SSCs important to safety that are designed and constructed using data

and/or analyses derived from ESP activities would perform satisfactorily in service. For example,

activities associated with data collection, analysis, and evaluation for soil composition, geology,

hydrology, and seismology determinations should be subjected to QA controls, commensurate with

the importance of the respective activities to design, and equivalent to the controls described in

Appendix B to 10 CFR Part 50. Further, information derived from recognized authorities, such as

the Census Bureau or the National Oceanic and Atmospheric Administration, should be controlled

using processes for data integrity, data traceability, document control, data- evaluation, data

analysis, and record storage that are equivalent to the processes and controls described in 10 CFR

Part 50, Appendix B.

The pre-application review places particular emphasis on the areas of organization, the QA

program, document control, and methodologies for data collection, analysis, and evaluation; It is

recognized that certain aspects of the applicant's quality controls may not fully implement the

18 criteria of 10 CFR Part 50, Appendix B, because not all criteria may be applicable to ESP

activities. However, the application should provide an adequate basis for evaluation of the

acceptability of the information in the application.

3.2 IMC-2502.."Construction Inspection Program: Pre-Combined License (Pre-COL)

Phase"

3.2.1 Introduction

This portion of the framework discusses the program of inspections necessary to support the NRC

staff's'preparation for a mandatory hearing before the ASLB and the final Commission decision on

whether a combined license may be granted. This support includes inputs to the safety evaluation

report (SER) and public meetings as necessary.

3.2.2 Inspections

The inspections governed by IMC-2502 will be implemented when the NRC is notified in writing that

a prospective applicant is preparing to apply for a COL. An application for a COL may, but isfnot

required to, reference a standard design certification, or an early site permit, or both. Therefore,

a COL application could include any of the following combinations10
- A standard design certification and an early site permit

- An early site permit and no design certification

- A standard design certification and no early site permit

- No design certification and no early site permit

The guidance contained in IMC-2502 will be flexible to accommodate these various options.

3.2.2.1 Quality Assurance Inspections During the Preparation of the Application

In the past, the NRC conducted meetings with prospective applicants and inspected their activities

during the preparation of license applications. The meetings and inspections were to provide

assurance to the reviewers about the quality of submittals. The NRC expects to continue this

'practice through the review of quality controls and the inspection of implementation of those

controls.

3.2.2.2 Engineering Design Verifications and First-of-a-Kind Engineering Inspections

In the past, NRC conducted design verification inspections late in the construction process. For the

next generation of plants, NRC plans to conduct independent design inspections as early in the

process as practical; however, these inspections may continue after the COL is issued. Inspections

to-support the decision to issue the COL will be conducted under IMC-2502 and will assess the

viability and implementation of, and results produced by, the applicant's design engineering

process. These inspections will assess the applicant's QA design controls and sample design

activities related to the site-specific portions of the plant's design.

Additional programmatic inspections, i.e., to monitor the design change process, may continue

under IMC-2504 after the COL is issued.

An inspection program will be developed for the inspection of first-of-a-kind (FOAK) engineering for

the lead plant of each certified design. The program will be described in a Inspection Manual

Chapter and will be similar to IMC-2530, "Integrated Design Inspection Program."

IMC-2502 reflects the following information on engineering design verification provided to the

Commission in SECY-94-294, "Construction Inspection and ITAAC Verification."

* Design descriptions and functional system drawings available for review during

the design certification and COL application stages are adequate for licensing

reviews and final safety determinations, but not for actual construction or

construction inspection activities.

* The NRC will inspect and review the adequacy of licensee design engineering

early in a construction project, possibly beginning soon after receipt of a

licensing application; first-of-a-kind engineering for the lead plant of each

certified design will be assessed during these inspections.

11
* NRC will also assess the effectiveness of the licensee's design change process

in maintaining the fidelity of high-level certified design information that is

translated into construction drawings.

3.2.2.3 Operational Program Reviews and Inspections

In SECY-02-0067, 'Inspection, Tests, Analyses, and Acceptance Criteria (ITAAC) for Operational

Programs (Programmatic ITAAC)," the staff provided its recommendation that COLs contain ITAAC

for operational programs required by regulations, such as training and emergency planning

programs. In response, the Commission's staff requirements memorandum dated September 11,

2002, directed the staff to develop guidelines regarding ITAACforoperational programs and towork

with stakeholders to resolve issues associated with the guidelines. The staff provided a response

to the SRM in SECY-04-0032, "Programmatic Information Needed for Approval of a COL Without

ITAAC." The staff position in the SECY paper involves developing COL review guidance and not

inspection guidance.

However, the staff's proposal in SECY-04-0032 discusses the possibility of an applicant submitting

implementing procedures at th COL stage to avoid programmatic ITAAC. For operational programs

which are submitted as part of the COL application, the NRC staff will conduct its first evaluation by

reviewing the bases of and inspecting any implementation of these programs. These inspection

activities will be conducted to support the review of the application under IMC-2502. If a program

has ITAAC (e.g., emergency planning) inspections will be performed under IMC-2503 to verify the

ITAAC. Additionally, SECY-04-0032 recognizes the possibility that inspections of programs that do

not have ITAAC will be done after a COL is issued. In such cases, inspections of these programs

will be done under IMC-2504.

3.2.2.4 Inspections of Other Activities Completed During the Pre-COL Phase

Changes being considered for 10 CFR Part 52 may allow for partial completion of some ITAAC prior

to the issuance of the COL. An example of an ITAAC that could be'completed during the licensing

review is the ITAAC for control room design, which includes the Design Acceptance Criteria (DAC)

for the control room without providing the engineering details. In such cases, the COL would

contain ITAAC to verify that the control room has been constructed in accordance with the design.

Two authorizations permitted under § 50.10(e) of the Code of Federal Regulations are informally

termed LWAs. These authorizations would also be permitted under § 52.91.

LWA-1: Under § 50.10(e)(1), the Director of Nuclear Reactor Regulation may authorize site

preparation work, installation of temporary construction support facilities, excavation for nuclear and

non-nuclear facilities, construction of service facilities, and construction of structures, systems, and

components which do not prevent or mitigate the consequences of postulated accidents. Sections

52.91 (a)(1) and-52.91(a)(2) contain the requirements for permitting LWA-1 activities for a COL.

Section 52.25 contain the requirements for permitting LWA-1 activities for an ESP. '

12
LWA-2: Under § 50.10(e)(3),-the Director of Nuclear Reactor Regulation may authorize the

installation of structural foundations for structures, systems,'and components which prevent or

mitigate the consequences of postulated accidents. An LWA-2 may be granted if, in addition to the

findings described above for an LWA-1, the ASLB determines that there are no unresolved safety

issues relating to the work to be authorized that would constitute good cause for withholding

authorization.

3.2.3 Construction Inspection Program Information Management System (CIPIMS)

The Construction Inspection Program Information Management System (CIPIMS) should be

available for scheduling and recording inspections necessary to support the application review.

This will be especially important for the documentation of information related to quality assurance

and engineering inspections, LWA activities, and ITAAC completions.

3.2.4 Enforcement

Enforcement actions associated with the application are not anticipated, but are not precluded,

during the COL review. However, the information submitted with the application will be subject to

NRC regulations, including -enforcement actions, for incomplete or inaccurate information. In

addition, the Commission has proposed to amend Part 52 to add requirements governing the

completeness and accuracy of information submitted by an applicant.

An early site permit referenced by the application for a COL is a license similar to a construction

- permit issued under 10 CFR Part 50. Therefore, violations of conditions of these licenses during

engineering design or LWA activities will be subject to enforcement, including notices of violation,

civil penalties and orders.

3.3 IMC-2503. "Construction Inspection Program: Inspections, Tests. Analyses, and

Acceptance Criteria"

3.3.1 Introduction

The Commission is required by § 52.97(b)(1 ) to identify within the combined license the inspections,

tests, and analyses that the licensee shall perform, and the acceptance criteria that, if met, are

necessary and sufficient to provide reasonable assurance that the facility has been constructed and

will be operated in conformity with the license, the provisions of the Atomic Energy Act, and the

Commission's rules and regulations. This portion of the inspection program framework document

discusses the inspection process used for the NRC's verification of licensee conclusions that the

ITAAC of a combined license have been met.

The results of this inspection program will provide input for the Commission's determination, in

accordance with § 52.103(g), of whether the ITAAC have been met and whether the licensee is

13
allowed to load fuel. The inspections to verify that ITAAC have been met should begin at the start

of placement of contracts for major component and module manufacturing and will end with the

Commission's decision on initial fuel load. Inspections of major components may occur before the

licensee receives its COL.

3.3.2 ITAAC Inspection Overview

Figure 3.1, 'IMCC-2503, Inspection Flow Diagram," depicts the basics of the inspection process for

this phase and the various types of inspection findings/assessments that will be made. Individual

ITAAC determinations of acceptability are made after inspection by NRC inspectors and

documented in inspection reports, as discussed below. It should be noted that there is only one

overall ITAAC conclusion made by the Commission relative to § 52.103(g). To support,the

§52.103(g) finding, the staff intends to use a phased verification method. This phased verification

method includes the concept of planning the inspection effort at the beginning of the process, using

a sign-as-you-go (SAYGO) throughout construction, publishing the results of NRC staff

determinations with respect to individual ITAAC as they are completed, and finally, making a

recommendation to the responsible regional administrator regarding the completion of all the

ITAAC.

The NRC staff expects that when an ITAAC is complete, the licensee will provide an ITAAC

determination letter that will demonstrate the satisfactory completion of the smallest increment of

an ITAAC. Examples are shown in an NEI letter to the NRC dated November 20, 2001 (ADAMS

accession number ML 020070338) and include, UITAAC 2.4.2 Item 7 High Pressure Core Flooder

System Remote Shutdown System Display," or ITAAC 2.4.2.2, 'High Pressure Core Flooder

System Hydrostatic Test." However, there are various possibilities of ITAAC determination letters

for the identification of ITAAC completion including:, an entire table (e.g., Table 2.1.1.d - Reactor

Pressure Vessel); a complete line item from a table (e.g., Table 2.1.2 -Nuclear Boiler System, Items

9.a. and 9.b.); or an individual item (e.g., Table 2.1.2, Item 9.a.). It is important that the licensee

schedule the performance of each ITAAC and communicate the schedule to the NRC in order to

allow NRC inspectors the opportunity to witness the performance of the ITAAC.

The master NRC construction inspection schedule will be derived from CIPIMS. CIPIMS will contain

all the ITAAC, the licensee's construction schedule and all related inspection resource information.

This system will also integrate all inspection results and correlate them with ITAAC and non-ITAAC

requirements and acceptance criteria. It will facilitate inspection at remote locations as licensees

make use of modular construction techniques, and integrate the inspection planning process with

the licensee's detailed construction scheduling process. It will allow the NRC to integrate the results

of inspection findings, ITAAC determinations, and the-staffs evaluation of licensee quality

assurance/quality control (QA/QC) effectiveness.

The ITAAC inspections will lead to two possible results: (1) inspection findings, and (2) ITAAC

determinations. An- example of an inspection finding would be documentation. of the, staffs

evaluation of the acceptability of licensee work processes that affect multiple ITAAC. A second

example of an inspection finding would be documentation of the staffs evaluation of a component

associated with a particular ITAAC. All inspection results will be documented in inspection reports.

14
No

/Information to make\ an ITAAC

e tion?

Yes

- - - I 4 -! I - Figure 3.1 IMC- 2503

:.= Inspection Flow Diagram

15
An ITAAC determination will document the staffs position on whether or not the licensee has

satisfactorily demonstrated that a particular ITAAC has been met (i.e., individual items in the ITAAC

tables). The regional administrator will be informed periodically on the status of ITAAC inspections

and all ITAAC determinations.

3.3.3 ITAAC Inspection Philosophy

Because the staff does not have the resources to perform direct inspection of all elements of all

ITAAC, the NRC will perform sampling-type inspections to verify that the licensee is in compliance

with NRC regulations. A combination of ITAAC sample selection, statistical methods, insights from

the probabilistic risk assessment (PRA), and inspections of the licensee's quality assurance

program will be used to help determine the necessary level of inspection effort and where limited

inspection resources are best spent. Appendix C of this document contains a general discussion

on inspection sampling. Work to establish a methodology for selecting appropriate inspection

samples was still going on at the time this document was issued as final. The specific guidance and

information about any methodology developed will be published at a later date.

The inspection program will rely on the licensee to ensure that all of the ITAAC have been met and

the inspectors will perform sampling type inspections to verify that the licensee has completed the

ITAAC in an acceptable manner. This will provide reasonable assurance that the facility has been

built and will operate in accordance with the license and the applicable regulations. The sampling

type inspections will be planned by the staff at the earliest stages of construction based on a review

of the ITAAC for the plant to be constructed. Because several ITAAC are expected to be closely

related, the staff may use the results of inspections for one ITAAC and apply them to other related

ITAAC. However, the staffdoes intend to perform a minimum set of inspections forall of the ITAAC.

The minimum set of inspections for all of the ITAAC is based on NEI's proposed process, set forth

in a November 20, 2001, letter, for informing the staff when an ITAAC or portion of an ITAAC is

completed. In accordance with this proposed process, the NRC staff expects that a licensee will

provide an ITAAC determination letter when ITAAC are completed. This letter will also inform the

staff that the bases for the determination are available for audit at the plant site. For those ITAAC

which have not received direct NRC inspection or a similar ITAAC was not inspected as discussed

previously, the inspectors will determine, at a minimum, if the licensee's ITAAC determination letter

and its associated bases are satisfactory by reviewing the documentation. The process that was

developed and used for certifying new reactor designs provided a risk-informed approach for

determining the ITAAC. Therefore, inspections conducted to verify satisfactory completion of ITAAC

provide a risk-informed inspection approach to the construction of new reactors.

3.3.4 ITAAC Inspection Process

3.3.4.1 Inspection Results

Inspections of ITAAC-related activities will be conducted in accordance with the inspection

procedures listed in IMC-2503. Inspection results will be documented in accordance with IMC-0613,

"Power Reactor Construction Inspection Reports." The staff has not yet developed IMC-0613, but

intends to do so within the next two years.

16
3.3.4.2 Review of Inspection Results

The review of inspection results will focus on two things: (1) the implementation of specific activities

as documented in the inspection history and (2) the implementation of the licensee quality controls.

The review would ensure that any deficiencies that have been-identified by the licensee or NRC

have been adequately addressed by the licensee's QA program and have resulted in effective

corrective actions. This would provide the NRC with confidence in relying on the licensee's quality

assurance program in assuring quality construction activities.

In keeping with the "sign-as-you-go" (SAYGO) approach, NRC staff and management will

periodically review inspection results to determine if the inspection history shows that sufficient

progress has been made in a specific area to reach an overall determination of acceptability., Based

on the inspection history, the staff could "sign-off' on the activities that have been found acceptable

and the level of inspection effort could be adjusted.

(1) Positive SAYGO Determinations

Should a review of the inspection history identify that activities are being effectively implemented

and deficiencies are being appropriately addressed, NRC staff will document their review as a

positive SAYGO determination. The inspection efforts associated with the particular construction

activity orwith a specific process may be reduced based on a positive SAYGO determination. This

determination could also reduce the inspection effort in other areas which are affected by this same

activity or process. The positive determination could also be used by the staff at a later time when

making the determination that ITAAC have been met by the licensee.

(2) Negative SAYGO Determinations

If the review of the inspection history identifies that a construction activity is not being effectively

implemented and that significant deficiencies are not being identified and appropriately corrected,

it could call into question the effectiveness of the licensee's quality assurance program and, if not

corrected, prevent the staff from making a positive ITAAC determination. NRC staff will document

their review as a negative SAYGO determination.

A negative SAYGO determination would call for the licensee to identify correction actions taken or

planned to address the specific identified deficiencies as well as the deficiencies in the corrective

action program. Further, the licensee would be expected to determine how the deficiency occurred,

whether or not it was generic, and to take actions to determine the extent of the condition.

NRC would consider increasing its inspection effort in this area by expanding the inspection sample

size to verify the extent of the condition and, if appropriate, re-examining other ITAAC which may

have the same or similar deficiencies. This could also increase the inspection effort in other areas

which are affected by this same activity or process. The NRC will verify the effectiveness of any

corrective actions. Upon verification of effective corrective actions, NRC staff would reassess the

construction activity, process, or component.

17
All reviews of inspection results would be documented in inspection reports and also reflected in

the CIPIMS database.

Table 3.1, "Examples of Construction Processes Appropriate for Evaluation Using a SAYGO

Approach," contains examples of construction processes which might be candidates for evaluation

using SAYGO. The list is intended to be representative rather than all-inclusive. Other construction

processes may also be appropriate and actual construction approaches may make the use of a

SAYGO approach inappropriate for some processes at some sites.

If, subsequent to the identification and documentation of a review of inspection findings, the NRC

determines that the results are-no longer valid (e.g., the NRC, licensee, or any other person

identifies new and significant information that has not been adequately addressed by the licensee's

corrective action program), the determination would be reassessed by cognizant NRC

management, communicated to the licensee, and documented in an NRC inspection report.

Consistent with past practices, the licensee would be afforded an opportunity to provide any new

information to the NRC which might affect the reversal of the previous review determination.

Table 3.1, Examples of Construction Processes

Appropriate for Evaluation Using a SAYGO Approach

Site preparation Concrete expansion anchors Heating, ventilation and air

conditioning

Mechanical penetrations Structural steel and supports, Conduit/tray supports

Equipment fabrication Safety related piping Conduit installation

Geotech/foundations Pipe support and restraints Tray installation

Structural concrete Welding Cable pulling

placement

Re-bar installation Masonry Cable terminations

Instrument sensing line Mechanical component/ Electrical component/

installation and piping equipment installation equipment installation

Nondestructive examination Electrical penetrations

3.3.4.3 ITAAC Determninations

As specific construction activities are completed, the licensee will determine that one or more ITAAC

have been completed. The licensee will document the specific inspections, tests, or analyses relied

upon in making the determination that one or more ITAAC are complete and ready for NRC

verification. This will be communicated to the NRC in the form of an ITAAC determination letter

requesting that the NRC staff verify that the ITAAC have been satisfactorily completed.

18
(1) Determining ITAAC Acceptability

Upon receipt of an ITAAC determination letter, the NRC will review the licensee's ITAAC

documentation and any NRC inspection reports related to that ITAAC. An NRC staff decision on

ITAAC acceptability will be called an ITAAC determination. The NRC's determination of ITAAC

acceptability will be based primarily on prior day-to-day onsite and offsite inspection activities,

interactions with licensee personnel, and inspection of construction activities in the field. These

inspections will have been documented in inspection reports.

In accordance with 10 CFR 52.99, the NRC will document each ITAAC determination in aFederal

Register Notice and in docketed correspondence to the licensee. The basis for determining the

acceptability of ITAAC will be documented in inspection reports and tracked in CIPIMS.

(2) Invalidation of Previously Accepted ITAAC Determinations

If new and significant information questions the validity of a previously accepted ITAAC

determination, the NRC would assess the information and determine the appropriate course of

action. The threshold that the NRC will use to determine what constitutes 'new and significant

information' that would invalidate, a previous ITAAC'determination is illustrated by examples in

Appendix D of this framework document.

Consistent with past practices, the licensee would be afforded an opportunity to provide any new

information (potentially including extensive corrective actions) to the NRC which might affect the

reversal of a previously accepted ITAAC determination. This information would be expected to

address whether or not the deficiency could be generic to other ITAAC and also why the extent of

the condition is or is not limited to this particular ITAAC. In addition, the licensee would be expected

to identify and correct the weaknesses in its corrective action program that allowed the deficiency

to occur. The NRC staffs decision on whether the ITAAC has been met would be communicated

to the licensee and would be made publicly available via the Federal Register.

3.3.4.4 Commission 10 CFR 52.103(g) ITAAC Finding

Before a facility may operate, the Commission is required by § 52.103(g) to find that the acceptance

criteria in the COL were met. Once the licensee has informed the staffthat all the ITAAC have been

completed, the staff will perform a review to ensure that an ITAAC determination letter has been

received for each ITAAC, a notice has been published in theFederalRegisterfor each ITAAC that

the staff has accepted, and the staff agrees that all the ITAAC have been met. The RA will rely on

the inspection and ITAAC determination results when informing the Director of Nuclear Reactor

Regulation (NRR) that all the ITAAC have -been met. The Director of NRR will make a

recommendation to the Commission that the Commission find that all acceptance criteria in the COL

have been met.

19
3.3.4.5 Public Notifications

The requirements for public notification of ITAAC completion are contained in § 52.99. -The staff

intends to publish ITAAC determinations in the Federal Register.: In addition, NRC inspection

reports, and correspondence with the licensee will be; published and be made available to the

public. The staff is considering the use of the NRC web site, similar to how it is used for the reactor

oversight program (ROP), as the chief electronic medium through which the results of inspection

activities can be made more readily available to the public.

3.3.4.6 Enforcement

During the construction period, the agency will process identified violations of NRC regulations and

conditions of the COL as set forth in the Commission's Enforcement Policy, NUREG-1 600, "General

Statement of Policy and Procedures for NRC Enforcement Actions," and will track these violations

in CIPIMS.

3.3.5 Inspecting Module Construction Activities

The use of a modular construction concept may be necessary to support the ambitious schedules

currently being proposed for the construction of the new generation of nuclear power plants. Offsite

fabrication of plant modules and plant components could begin well before COL issuance. There

may be some instances where a basically complete plant is fabricated at one offsite facility. In such

cases, the requirements to have a manufacturing license in accordance with 10 CFR Part 52 may

apply.

Major plant components such as reactor pressure vessels, steam generators, and reactor coolant

pumps, as well as smaller components such as electrical breakers, relays, and valves,; have

traditionally been fabricated at an offsite location. For future nuclear power plants, large portions

of the plant could be modular in design, allowing for offsite fabrication' and assembly of portions of

buildings and rooms containing completed and tested systems and subsystems.

Discussions with several design and construction organizations concerning modular construction

have convinced the staff that as much as 60 percent of what had been site construction activities

in the past will probably be moved offsite to the locations where the modules will be fabricated.

NRC oversight activities in addition to those performed during the construction of the existing fleet

of nuclear plants will be necessary to assure that an acceptable level of quality is maintained

throughout the fabrication or manufacturing process. Collectively, these offsite construction

activities pose significant challenges for the planning and implementation of NRC inspections.

One example of such a challenge arose during the construction of the ABWR in Taiwan. On that

project, a problem occurred during the fabrication of the reactor pressure vessel pedestal. The

problem involved offsite fabrication as well as onsite construction by the fabrication contractor. The

staff believes that this problem provides valuable "lessons leamed" for the inspection of fabrication

facilities and remote process activities. Appendix E to this document contains information regarding

the problem.

20
For all offsite manufacturing and fabrication activities, a major focus of NRC review will be assuring

acceptable licensee QA oversight. In addition, for ITAAC-related components and modules, the

NRC may perform inspections at the offsite location. A major focus of these offsite location

inspections will be assuring that the vendor has implemented QA requirements appropriately.

Additionally, the staff intends to inspect the programmatic implementation of QA requirements at

vendor facilities and will assess the overall effectiveness of vendor QA activities under IMC-2504.

Appendix C of this document provides a more detailed discussion of the role of the quality

assurance program as it related to ITAAC. As discussed in Appendix C, QA deficiencies may

impact the NRC's ITAAC determinations.

The NRC would expect to use the above-mentioned phased verification process and apply it to

offsite inspections where appropriate. For example, if safety-related pipe welding is taking place

in several different offsite fabrication facilities (e.g., shipyards) because of modular construction, and

also onsite (to connect one safety-related module to another), then inspection findings and

assessments could be used for the individual offsite fabrication facilities as well as for onsite

activities.

Similarly, an inspection finding or assessment could be used at manufacturing facilities that are

supplying the reactor pressure vessel, steam generators, etc. If a positive SAYGO determination

is documented, site inspections could be limited to inspections for handling and shipping damage

after the component arrives at the site.

Figure 3.2, uAnticipated Nuclear Power Plant Construction Schedule," lays out a typical construction

schedule for a nuclear power plant using modular construction techniques. A gas-cooled reactor

vendor indicated that the time from first placement of structural concrete to fuel load (all ITAAC met)

was projected to be approximately 20 months per module, while another light water reactor

applicant indicated that the construction time frame for its design would be 42 months. The 36-

month time frame in the figure is, therefore, meant to be representative and the schedules will be

different based on the design and the applicant. The top of the figure shows the applicant's

schedule and at the bottom of the figure are the staff's two high-level IMCs that will guide the

inspection activities associated with this schedule.

The time line for IMC-2503 shows the major milestones associated with the staffs ITAAC inspection

activities. These inspection activities would start with inspections associated with an overview of

the applicant's QA program and how the applicant will satisfy the QA requirements associated with

the fabrication of the major components and modules. The process ends with the Commission

decision regarding ITAAC. This time line shows when the staff expects to perform inspections that

will have a direct effect on ITAAC determinations. With the heavy reliance on modular construction,

the staff fully expects to be performing inspections both onsite and at offsite facilities to support

ITAAC determinations.

21
Re-Co~kcin Cbristnadon arWI

18 Moriff I 36 Marith

9 Urmbi VVbA~di

0 ln re

I~l~s

I bA Trtks

15cavabo o Frstuduh Qxre

rBoWn

AF1mJ 11

I

I

I

I

I

I

I

I

I

I

I

iA-MnI

0ilLa 0

MduRrg g I 

SetRnay quientHo

Tubre

FRN Hearing Opportunity (not Issj Staff recommendation ti,-, I Ft g -j ,,- f. %l i-n.4 Am A A t . n. ,

set

Ger

I

I

I

IMC-2503 RC ITMC Inspections begin

begins A9

IMC-2504 ANRC Non-TMC Inspedions

begins

;OL Issued Susan orb ITAAC rams baa . ...... --.. -- r-.s

Operatonal Red lowed i Commisson finds ITMC met)

AA-- -AIMC-2515 ; .

Figure 3.2 - Anticipated Nuclear Power Plant Construction Schedule

The second time line is for IMC-2504, which is discussed in more detail in Section 3.4 of this

document. This time line is meant to illustrate the staffs inspections associated with things other

than ITAAC, the transition to IMC-2515, and the reactor oversight program.

3.4 IMC-2504. "Construction Inspection Program: Non-ITAAC Inspections"

3.4.1 Introduction

The purpose of this section of the framework document is to discuss the inspections, other than

those associated with ITAAC, that the NRC expects to perform from the time that a COL is issued

(or earlier if necessary) until sometime after the plant reaches full power operations status, i.e.,

when IMC-2515 begins. Inspection guidance for a plant licensed in accordance with Part 50 that

is relevant to a plant licensed in accordance with Part 52 is contained in IMC-251 and IMC-2514.

The scope of inspections will include structures, systems and components that are safety related

to ensure that the licensee is in compliance with appropriate rules. In addition, other issues that are

not safety related but are considered important to safety, such as station black out, will also be

inspected.

22
Because of the different issues involved with this period, the time frame is broken into two

segments; the time period before the § 52.103(g) finding is made by the Commission, and the time

period from the §52.103(g) finding until the point when IMC-2515 begins. This division is needed

because, at the time of the §52.103(g) Commission finding, several things occur that affect the

inspection program. For example, ITAAC end, while emergency planning, and technical

specifications requirements begin. Inspections associated with each time period are discussed in

the sections below.

3.4.2 Before 10 CFR 52.103(g) Finding

If an operational program has an ITAAC associated with it, then IMC-2503, discussed above, would

govern inspections associated with that program. The staff will perform inspections of programs

early in construction and prior to operation to verify the licensee's compliance with regulations. The

Commission's SRM associated with programmatic ITAAC, dated September 11, 2002, recognized

that because not all operational programs will have an ITAAC, the staff can take appropriate

enforcement action to prohibit or delay fuel loading pending appropriate corrective action if the

licensee's operational programs do not provide adequate protection of public health and safety.

Inspections of programs that do not have an ITAAC would be done similarly to the construction

inspection under 10 CFR Part 50. The inspection guidance for operational programs is contained

in existing inspection procedures which will be implemented by IMC-2504. The guidance previously

contained in IMC-2513 for pre-operational testing under 10 CFR Part 50 will be edited and

incorporated into IMC-2504. The staff expects to augment such inspection guidance with lessons

learned from the ROP. This may include using some of the inspection procedures usually

implemented under IMC-2515 to assess some operational programs and doing inspections in the

cornerstone inspection areas identified in the ROP. The cornerstone inspection areas are reactor

safety, radiation safety, and safeguards, while the cross cutting elements are human performance,

safety conscious work environment, and the corrective action program. Inspections done in these

areas are separate from the ITAAC determinations. To the extent that they are performed prior to

loading fuel, these inspections will supplement the bases for the regional administrator's

recommendation to the Director of NRR regarding the licensee's overall readiness to load fuel.

Engineering design inspections will continue as the licensee completes site-specific designs to

address the design acceptance criteria (DAC) and to document the final as-built system

configuration.

3.4.2.1 Inspection After an ITAAC is Met and Prior to Fuel Loading

There is guidance in both draft SRP 14.3 and in SECY papers to the Commission on performing

inspections after an ITAAC determination is made but before the Commission makes its

23
determination in accordance with § 52.103(g). Draft SRP section 14.3, uITAAC Design

Certification," provides some guidance in this area. The following is an excerpt from Section 14.3,

Appendix A, Section IV.B.2 of the SRP:

The purpose of the ITAAC is to verify that an as-built facility conforms to the

approved plant design and applicable regulations. When coupled in a COL with the

ITAAC for site-specific portions of the design, they constitute'the verification

activities for a facility that must be successfully met prior to fuel load. If the licensee

demonstrates that the ITAAC are met and the staff agrees that they are successfully

met, then the licensee will be permitted to load fuel. Once completion of ITAAC and

the supporting design information demonstrate that the facility has been properly

constructed; it then becomes the function of existing programs such as the technical

specifications, the in-service inspection and in-service testing program; the QA

program, and the maintenance program, to demonstrate that the facility continues

to operate in accordance with the certified design and the license.

The staff has- also provided guidance; on when programs such as technical specifications are

required to be in place. Most recently, the proposed rule for Part 52 contains a revised section on

applicability of NRC requirements (see section 52.215(c) on page 122 of the attachment to SECY02-0077). This section makes it clear that the COL holder does not need to be in conformance with

most operational requirements until the Commission has authorized fuel load. The staff also

proposed in SECY-00-0092, " Combined License Review Process," license condition 2.1 (see

Appendix F) which states:

The following operational requirements that are applicable to this license will become

effective after the Commission finds that the acceptance criteria in this license (COL

ITAAC) have been met in accordance with 10 CFR 52.103(g):

* emergency plans

* technical specifications

Therefore, there is the expectation that a license condition will specifically designate when some

operational programs will be required to be in place.

Because of the above, the staff recognizes that there is a period of time between when an individual

ITAAC is completed and when the Commission's § 52.103(g) finding is made. Thus, there will be

a period of time from completion of an individual ITAAC to when the programs, such as technical

specifications, are required to be in place to demonstrate that the facility continues to operate in

accordance with the certified design and license. During this period of time, the NRC will perform

inspections in accordance with IMC-2503 to verify that the ITAAC determinations remain valid.

3.4.2.2 Operational Readiness Assessments

In the past, an operational readiness inspection was done in accordance with Inspection Procedure

93806, "Operational Readiness Assessment Team Inspection (ORAT)," to ensure that a plant was

prepared for its low power license. Operational readiness inspections will continue to be performed

24
for plants licensed in accordance with 10 CFR Part 52. The results of the inspections could provide

significant information separat from the ITAC completion status to the regional administrator and

Director of NRR regarding the licensee's operational capability and organizational readiness to load

fuel.

Inspections to observe the development of programs necessary to support major activities in

construction and turnoverto operation are expected to begin early in the construction process. The

staff expects the licensee to phase in the operational programs necessary to support each

milestone of construction before the program is required by regulations. These inspections would

verify that the licensee is ready to implement programs such as licensed operator training, security

and fire protection, which would be necessary to support fuel load. The staff expects to inform the

Commission of the status of these programs before a Commission decision is made relative to

§52.103(g).

3.4.3 Post Fuel Load Prior to Power Operations

As discussed earlier, ITAAC end when the Commission makes the findings required under § 52.103

before operation. Therefore, after fuel load, the ITAAC do not constitute regulatory requirements

for the COL holder. Adequate protection of the public health and safety during plant operation is

assured by continuing compliance with the terms of the COL, including technical specifications, and

the NRC's regulations. The inspection guidance for this phase of the construction inspection

program will be contained in IMC-2504. In addition, as discussed abovejlessons learned from the

ROP will be implemented so that the transition to IMC-2515 is smooth. In general, IMC-2504 will

support the remainder of the pre-operational testing (e.g., integrated hot functional test and loss of

offsite power), the startup testing (pre-criticality tests, low power physics testing, and power

ascension testing), and transition to IMC-2515.

The staff will continue to perform inspections to support major milestones after fuel loading and

before IMC-2515 begins. Such milestones include low power and full power operations. The staff

expects that the RA will inform the Director of NRR on the licensee's readiness to achieve these

milestones.

3.5 Construction Inspection Program Information Management System

(CIPIMS)

CIPIMS is a dedicated, computer-based inspection scheduling and information management system

intended for deployment at nuclear power plants (NPPs) under construction. CIPIMS will be used

to organize and manage inspection information and will integrate the licensee's construction

schedule, inspection results and findings to support ITAAC determinations. CIPIMS will provide

a standard, consistent, systems-based apprbach to coordinating, scheduling, collecting, organizing,

and recording inspection data necessary to establish a reasonable assurance finding for ITAAC

determinations and eventual transition activities from construction to operational inspection under

IMC-2515.

25
The CIP team and NRR's Work Planning Center (WPC) initiated an effort to adapt the CIPIMS

software described in Attachment 4 of the "Draft Report on the Revised Construction Inspection

Program," dated October 1996, to the CIP provisions of today; In addition to updating the software,

the team also faced the challenges of interfacing the system with the Agencywide Documents

Access and Management System (ADAMS), which was created after the original CIPIMS was

developed, and applying new technologies that will support CIP efforts to economize inspection

resources.

The combined efforts of the CIP team and NRR's WPC resulted in redefining the needs to be met

by CIPIMS to accommodate future construction techniques, quality processes, inspection program

management, skills and experience, inspection program structure and implementation, inspection

documentation, and inspection planning and activities.

CIPIMS will use:

* Smart coding

* Integrated scheduling

* Tablet personal computers

* Barcode technologies for data collection and tracking-

* Computer aided engineering (CAE) design tools

* Digital imaging

* Lessons learned from international programs and feedback on foreign parts fabrication

* CIPIMS should have native compatibility with licensee information systems and

technologies (licensee PrimaveraO P3 scheduling systems, Microsoft® Office, etc.).

The CIPIMS scheduling software should be able to easily interface with the licensee's scheduling

software. The vision for CIPIMS is that the NRC inspection scheduler, working with the respective

region and headquarters, will plan construction inspection activities in advance based on the initial

schedule from the licensee. The schedule would then be automatically updated as the licensee's

schedule changes.- The licensee may not know the inspection activities the NRC has planned in

advance but it is vitally important that the staff have up-to-date accurate information regarding the

construction schedule, including those activities that are being performed offsite.

A series of meetings with Westinghouse, General Electric, Atomic Energy of Canada Limited, and

Bechtel to discuss their construction scheduling software revealed that each is currently using

PrimaveraO as their scheduling software program. However, they acknowledge they may

eventually move to other products. The NRC has noted that the level of detail in some -master

schedules, where only the delivery date of a major component or module is identified, may make

scheduling an inspection more difficult.

The NRC and NEI have established a working group specifically to test CIPIMS. The group will

establish the level of detail, coding structure, and transfer protocols needed to efficiently transfer

schedule information for use in CIPIMS. These outcomes will be achieved by developing and

transferring detailed schedules on'selected sample work streams as a means of testing the various

attributes of CIPIMS.

CIPIMS will be successful only if the licensee, the prime, sub, and fabrication contractors, and the

NRC use a common coding schema allowing seamless integration of licensee schedules with

26
ITAAC procedures. The adoption of a standard coding schema will provide for auditing and

traceability of the inspection process at a level of detail previously not available to the NRC. It

should be noted that if fabrication occurs in geographically dispersed areas, support for universal

time and date formats will be necessary to ensure proper resource scheduling and availability.

3.6 NRC Organization

3.6.1 Introduction

Upon receipt of a written notice of intent to submit an application for an ESP or a COL, NRC

management will establish an NRC organization to implement the CIP for the proposed site.

3.6.2 Implementation

The CIP team determined that although the specific organizational composition would be best

defined at the time of implementation, there were some important factors to be considered when

initially establishing this NRC organization.

3.6.2.1 ESP

The area of inspection is fairly limited for an ESP application, and therefore major organizational

considerations are not involved. It is important that NRR and the Region identify points-of-contact

(i.e.,' a project manager or a' project engineer) as soon as an applicant announces the intent to

submit an ESP application. These points-of-contact are necessary for the coordination of preapplication site visits, review inspections, and public meetings.

3.6.2.2 COL and Beyond

Inspections to support the review of a COL application are more involved than inspections for an

ESP. Once again, it is important that NRR and the Region identify points-of-contact early (i.e., as

soon as the intent to submit a COL is announced) in order to coordinate necessary pre-application

activities. One important function of the pre-application contacts with an applicant is to gauge the

applicant's proposed construction schedule'and projected offsite construction plans, in order to

determine when the NRC should begin organizing its construction inspection team.

3.6.3 Items To Be Considered During the Development of the Inspection Organization

The licensing and construction environment needs to be considered when establishing the

organization. During discussions with design organizations, the CIP team was informed that by

using modular construction techniques, as much as 60 percent of what in the past were considered

27
typical onsite construction activities will be moved off site. It is also conceivable that some offsite

modular and component fabrication activities could begin prior to the submittals of the application

for a COL.

Depending on the extent of modular construction employed, the inspection staff may need to

conduct inspections at remote locations, such as U.S. or foreign shipyards or fabrication facilities.

Because of modular construction, overall site construction schedules will be significantly accelerated

compared to those of past construction projects. Site construction'duration from the first safetyrelated concrete placement until request for permission to load fuel could be as short as 24 to 30

months.

Another consideration which would have an impact on the development and location of an

inspection organization is whether one COL application will be submitted or whether multiple

applications are expected.

Considering these factors, the CIP team identified the following as necessary to ensure the success

of the initial organization.

3.6.3.1 Regional Involvement

The regional office should oversee the implementation of, and the overall coordination of, the

inspection program for a particular site. This is necessary to ensure that the cognizant regional

administrator is involved throughout the process so that the RA may make a recommendation to

the Director of NRR regarding the completion of ITAAC and other prerequisites for initial fuel load.

The regional office should oversee the onsite inspection program and would provide inspection

resources and other technical support as necessary.

3.6.3.2 Inspection Scheduling and Data Management Activities

There will be a critical need for a central scheduler. There is a likelihood that the CIP

implementation could be performed using one organization to address onsite construction and a

different organization to address remote fabrication of components and modules. Inspection results

from separate inspection locations could be used to complete a single ITAAC.

In order to plan for and to coordinate all of the required inspection activities, the scheduler would

be directly responsible for communicating with the applicant's scheduling organization and for

coordinating all inspection activities with the respective inspection team leaders. All inspection

activities coordinated through the central scheduler would need to be planned, scheduled, and

tracked through completion using a central data system such as CIPIMS.

The scheduler would be trained in a scheduling program that is compatible with the applicant's

scheduling software. This individual would also be trained on the use of the CIPIMS and would be

responsible for the overall utilization and maintenance of the CIPIMS data for that site.

28
-3.6.3.3 Inspection Project Management

In the earliest phases of plant construction, inspection activities could be governed by three different

inspection programs, IMC-2502, -2503 and -2504, during the same time frame. The inspection staff

needed for the planning of the inspections should start to be assembled just before, or at least no

later than, the docketing of the application and initiation of the licensing review.

The inspection planning function could operate from either the cognizant regional office or NRC

headquarters, and should be conducted by the inspectors and scheduler selected to staff the

resident inspectors' office at the site. This activity would be expected to shift to the site with the

advent of significant safety-related site construction activities needing inspection coverage.

As stated above, construction inspection activities will be performed off site, as well as on site, and

could involve inspections from at least three separate inspection programs. Therefore, there could

be two or more separate organizations involved with inspection implementation for a single project.

Remote inspections/audits of component and module fabrication activities, as well as design and

engineering activities, could be conducted by different implementing organizations that would

coordinate with the onsite construction inspection team using the CIPIMS. Examples of some of

these inspections are as follows:

* First-of-a-kind engineering inspections under IMC-2502

* Inspections of detailed design information provided in place of DAC for approved

designs (e.g., instrumentation and human factors of control room design), under IMC2502 (Note: The process for resolving DAC prior to issuance of a COL is being

considered as part of the 10 CFR Part 52 update rulemaking)

* ITAAC inspections under IMC-2503 for site construction activities

* - ITAAC inspections under IMC-2503 for remotely manufactured modules or components

* Inspection of programs under IMC-2504

* Special Inspection programs

Special inspections/audits and foreign manufacturer inspections/audits are expected to be

performed, or at least coordinated, out of headquarters.

An inspection program like the vendor inspection program would likely be needed for remote

inspection/audit activities. These activities would be coordinated by either the Director of the

Division of Inspection Program Management in NRR, by the lead region Division of Projects, or by

a regional 'center of excellence.'

- An inspection program,-similar to IMC-2530, Integrated Design Inspection Program, should be

developedforthe inspection of first-of-a-kind (FOAK)engineering forthe lead plant. Amanagement

decision would be needed to determine whether this FOAK engineering inspection would be

conducted, or led, by the primary Region for the lead plant, the Region in the geographic location

29
of the designer, a Region with a center of excellence in the area of design, or an NRC inspection

group located in headquarters.

The October 1996 uReport on the Revised Construction Inspection Program" provided valuable

insights into how an appropriate organization may look. In addition, the CIP team has developed

the proposed organization described in Appendix G, "Example of an NRC Organization for

Implementing a CIP," which parallels that of the Special Projects organizations of the late 1980s and

early 1990s for TVA and Comanche Peak. Upon notification of the pending submittal of a COL

application, these examples as well as the issues discussed above should be reviewed and

evaluated in the context of a contemporary NRC organization-to ensure that the CIP will be

effectively and efficiently implemented.

3.6.3.4 ITAAC Comparison to Inspection Procedures

As part of an effort to determine how much of the old construction inspection program can be

applied to future plants, the staff has reviewed the ITAAC that were developed for the ABWR and

the AP600. A similar effort was previously done for the ABWR in SECY-94-294 for the highpressure core flooder system. The staff broadened this effort to include all of the ITAAC for both

plants to identify what inspection procedures will be used for an ITAAC and begin to estimate the

amount of work involved in writing and rewriting the inspection procedures. Appendix H to this

framework document contains the results of this review.

3.6.6.5 Inspection Findings and Enforcement

All inspection findings identified during the new construction period will be documented in

accordance with IMC-0613 after they have been placed in context and, assessed for impact on

ITAAC. For enforcement purposes, the new construction period starts once the COL is approved

for the facility and ends when the unit enters poweroperations. Once a COL is applied for, some

regulations apply and enforcement actions may be considered for identified violations. During this

period, potential violations from inspection activities will be processed in accordance with 10 CFR

Part 2, the Enforcement Policy, and other applicable enforcement guidance using traditional

enforcement tools. The findings will then be categorized as violations, deviations, nonconformances, or unresolved items. This includes use of severity levels, notice of violations (NOVs)

for violations of severity level IlIl and above, and civil penalties as appropriate.

Once the facility enters power operations, there will be a transition to the reactor oversight process

(ROP). During this transition period, inspection findings and enforcement actions will be processed

using the ROP as much as practicable. The approach for transitioning to the ROP will be as follows:

The facility transition to the ROP will be a gradual-phased approach on an individualized

cornerstone basis. The basis for determining that a cornerstone is ready to be monitored under the

ROP will be documented. The document will contain all the records that verify that a cornerstone

can be monitored fully.

30
When one or several cornerstones appear to be ready to be monitored under the ROP, a transition

plan will be developed which will specify which inspections will be performed to verify that all issues

have been resolved and that all licensee corrective actions are effective.

The regulatory response and plant performance assessment will be in accordance with the Action

Matrix as defined in IMC-0305, "Operating Reactor Assessment Program." During the transition to

the ROP, the regulatory responses allowed by the Action Matrix may be used with the concurrence

of the management team assigned responsibility for construction and the regional administrator.

The transfer of the facility to the full reactor oversight process will be accomplished by written

approval of the regional administrator with the concurrence of the Director, Office of Nuclear

- Reactor Regulation (NRR). This transfer may occur even if all performance indicators are not yet

available, provided compensatory inspections are conducted as provided for by IMC-2515. The

management team assigned responsibility for construction may be dissolved at that time or may be

maintained for up to two additional quarters if necessary to deal appropriately with outstanding

issues.

31
4. PUBLIC INVOLVEMENT

Public involvement has been an important aspect of the development of this document. Every effort

has been made to seek input from both internal and external stakeholders 'during the process of

developing the scope and defining the content of the construction inspection program for plants that

might be built under I0 CFR Part 52. This document was reviewed by the Advisory Committee on

Reactor Safeguards (ACRS) in December 2003. The recommendations made by the ACRS

(ADAMS Accession Number ML033460266) have been incorporated.

Public comment on this document was sought through a Federal Register Notice in May 2003,

when the framework was initially proposed. A subsequent workshop in August 2003, offered the

opportunity for the NRC staff to provide more in-depth descriptions and explanations' of the'activities

planned under the CIP and detailed in the framework. The workshop also provided an opportunity

for external stakeholders'to ask questions and suggest alternatives for various aspects of the

framework. Ideas and issues raised during the workshop were captured in a written transcript of

the day-long meeting.

Written comments submitted in response to the Federal Register Notice as well as suggestions

and questions raised during the workshop were considered and incorporated to the extent possible

into this version of the framework. A detailed listing of the various comments as well as an

explanation of how the comment was resolved can be found in Appendix I, "Comment Resolution

Summary for the Draft Construction Inspection Program Framework Document."

32
Appendix A

Glossary

Attribute Guidance. The Inspection Guidance, generally discussed in each NRC inspection

procedure, that relates to the types of activities an inspector should observe and review and

together with some references, providing specific acceptance criteria that can be used in the

evaluation process.

Audit. An applicant/contractor activity to determine through investigation the adequacy of, and

adherence to, established procedures, instructions, specifications, codes, and other applicable

contractual and licensing requirements, and the effectiveness of implementation

Construction Inspection Program Information Management System (CIPIMS). CIPIMS is a

dedicated, computer-based inspection scheduling and information management system intended

f6r deployment at nuclear power plants (NPPs) under construction.' CIPIMS will be used to

integrated the inspection schedule with the licensee's construction schedule. It will also be used

to organize and manage information about the inspection results, the licensee's ITAAC completion

information, and the NRC's ITAAC determinations.

Combined License. A combined construction permit and operating license with conditions issued

pursuant to 10 CFR Part 52. Like a construction permit under 10 CFR Part 50, a combined license

under 10 CFR Part 52 authorizes construction of a nuclear power plant. The NRC ensures that the

licensee has completed the required inspections, tests, and analyses, and authorizes operation after

finding that the acceptance criteria have been met.

Construction Activities. Any activity associated with the construction, fabrication, or testing of

structures, components, subcomponents, subsystems, or systems either at the construction site or

at remote fabrication or testing facilities that occurs during the construction phase of the inspection

program. Construction activities also include the design and engineering of the structures, systems,

and components of the facility.

Contractor. Any organization under contract for furnishing items or services to an organization

operating under the requirements of Appendix B of 10 CFR 50 or the commitments made in the

application. It includes the terms Consultant, Vendor, Supplier, Fabricator, Constructor, and subtier

levels of these, where appropriate.

Critical Attribute. A characteristic or quality of any construction material, object, action, or process

that demonstrates that design and performance requirements have been met either uniquely forthe

item or collectively for the related structure, system, or component. Critical attributes (which may

also be applicable to construction documents' such as procedures, reports, and records) are

discussed in each NRC inspection procedure as Inspection Requirements, delineating specific

inspection activities that may be conducted to check the listed attributes for conformance with the

relevant acceptance criteria.

A-1
Design Control Document. The design control document is a repository of information on the

respective standard plant design (e.g., AP600, advanced boiling water reactor). The design control

document also provides the design-related information that is incorporated by reference into the

respective appendix to 10 CFR Part 52. The design control document consists of Tier 1 and Tier

2 information (see below for definitions).

Documentation. Any written or pictorial information describing, defining, specifying, reporting, or

certifying activities, requirements, procedures, or results.

Early Site Permit. Under 10 CFR Part 52, an early site permit addresses site suitability issues,

environmental protection issues, and plans for coping with emergencies, independent of the review

of a specific nuclear plant design.

Inspection. (1) An NRC activity consisting of examination, observation or measurements to

determine applicant/contractor conformance with requirements and/or standards. (2)

Applicant/contractor quality control measures consisting of -examination, observation- or

measurements; to determine the conformance of. materials, supplies, components, parts,

appurtenances, systems, processes or structures to pre-deter-mined quality requirements.

Inspection Finding. A documented evaluation of the acceptability of licensee construction activities.

Inspection Sample. An item selected for inspection of one or more critical attributes.' For example,

an inspection sample may be a single record for review of welding records, while an entire system

would comprise the inspection sample during a system walkdown inspection. The composition of

an inspection sample will be defined in each inspection procedure under the sampling criteria. The

inspection sample should be identified in CIPIMS with the licensee's unique identification number.

Inspections. Tests. Analyses, and Acceptance Criteria (ITAAC). ITAAC are a provision of the

Atomic Energy Act and 10 CFR Part 52. They are necessary to ensure that a plant licensed in

accordance with 10 CFR Part 52 has, been properly constructed and will operate safely. The

licensee performs the ITAAC and the Commission must find that the ITAAC have been met before

fuel loading at the nuclear power plant is allowed.

ITAAC Determination. A determination about the completion of an ITAAC that is made by the

inspection staff after reviewing the inspection history and the licensee's documentation related to

the ITAAC. This determination is performed for individual ITAAC and when combined for all ITAAC

will lead to a recommendation that the Commission makes a finding in accordance with 10 CFR

52.103(g).

Lead Region. The region designated with the authority to make a recommendation to the Director

of Nuclear Reactor, Regulation that an activity has' been satisfactorily completed based on

inspections associated with an early site permit or combined license application. The lead region

is based on geography and is defined as that region that oversees the location of an early site

permit or combined license.

Limited Work Authorization (LWA). Authorization from the NRC to an applicant to conduct certain

construction activities pursuant to 10 CFR 50.10(e)(1) or 10 CFR 50.1O(e)(3)(i).

A-2
Sian As You Go (SAYGO). Process. For selected construction activities, the NRC will perform

inspections beginning during the early stages of reactor construction to assure that construction

activities are accomplished in accordance with licensee procedures, applicable codes and

standards, and NRC regulations. In addition, the NRC will check to ensure that the licensee has

implemented QA/QC oversight of these activities such that acceptable quality is consistently

maintained. If the activities are resulting in consistently satisfactory results, the NRC will 'sign-off

on the activity and will consider reducing the inspection effort in that area.

Tier 1 Information. Tier 1 information is that portion of the design-related information in the design

control document that is approved and certified by the NRC through rulemaking. Tier 1 information

includes the following:

* definitions and general provisions

* design descriptions

* ITAAC

* significant site parameters

* significant interface requirements

Tier 2 Information. Tier 2 information is that portion of the design-related information in the design

control document that is approved but not certified by the design certification rule. Tier 2 information

includes the following:

* information required by 10 CFR 52.47, with the exception of generic technical specifications

and conceptual design information

* information required for a final safety analysis report under 10 CFR 50.34

* supporting information on the inspections, tests, and analyses that will be performed to

demonstrate that the acceptance criteria in the ITAAC have been met

* combined license information items which identify certain matters that shall be addressed

in the site-specific portion of the final safety analysis report by an applicant who references

a design certification rule

A-3
Appendix B

Information Considered in Updating

the Construction Inspection Program

- This appendix discusses lessons learned from the 1996 revised construction inspection program

document which was considered in the development of the 10 CFR Part 52 Construction Inspection

Program (CIP). The report covers a variety of programs, activities, and experiences from the last

- NRC construction inspections conducted at Seabrook, Comanche Peak, South Texas, Watts Bar,

and Bellefonte. In updating the CIP, the staff considered the previously reviewed foreign

construction inspection practices and the modular construction techniques used in the US

shipbuilding industry.

Quality Processes

* The assessment process must begin with inspections of the design engineering process,

- including engineering quality assurance (QA), to ensure that the licensee can accurately

translate high-level design requirements into detailed engineering and fabrication drawings.

* Because NRC inspections are done on a sampling basis, the CIP must provide accurate

assessment of the licensee's quality programs. To the extent possible, all construction

inspections should assess the effectiveness of QA and quality control (QC), and the results

must be thoroughly documented and integrated. Additionally, the staff intends to perform

programmatic QA inspections to provide reasonable assurance that Appendix B

requirements are adequately implemented. Ideally, the breadth and depth of the NRC's

verification that a plant's QA and QC are effective will ensure that any demonstrated or

alleged lapses in quality are isolated instances rather than generic problems.

* The licensee's management of QC records is an integral part of the quality process. In

order to verify the overall adequacy of the licensee's QA records management process, the

CIP must inspect all aspects of QA/QC records, from creation through storage.

* - The identification of construction problems and the timeliness and extent to which they are

corrected are effective measures of licensee management's control over onsite activities.

NRC experience shows.that, if the licensee has a thorough corrective action program and

effectively identifies and corrects root causes of problems, there is a good chance that the

overall quality of the construction is good. If the corrective action program is weak, it is likely

that there are lapses in quality (i.e., if repetitive problems occur).

Inspection Program Management

The objectives of the inspection program is to support the Commission's 52.103(g) finding that all

ITAAC have been met and that programs are in place to ensure the facility will operate in

conformance with the Commission's regulations. This approach will be more likely to produce

enough inspection data to assess the adequacy of a plant's construction and readiness to

commence operations. These- objectives should be considered in establishing the inspection

B-1
methodologies to be employed (e.g., risk-informing the inspection sample selection, inspection

sample size) and the format and content of inspection documentation.

In the past, NRC construction inspections were often scheduled on the basis of inspector

availability. Inspections were therefore performed on activities that happened to be in

progress at the time of the inspection, resultinrg in a less-than-optimum sample selection.

Because inspectors will be continuously onsite under the revised CdP, and because ITAAC

must be verified under Part 52, NRC inspections must be scheduled on the basis of the

utility's construction progress. All aspects of the construction inspection program, including

inspection planning, scheduling, preparations, and implementation, must be' conducted in

a way that will ensure'all necessary attributes are properly inspected.

* The proper mix of skills and experience among inspectors, particularly during the near-term

operating license (NTOL) phase at a plant, is necessary to ensure effective implementation

of the inspection program.

* The CIP must be able to support NRC action on a licensee's certification of readiness to

load fuel, all ITAAC having been completed satisfactorily. The inspection staff should be

fully aware, in advance, of all issues the licensee will address in its certification.

* Inspection results must be assessed to verify that inspection requirements are met and that

the results support the objectives of individual inspection procedures and of the CIP.

* A plan for the transition from the construction phase to the operations phase should be

made well in advance of the completion of 'plant construction. This transition plan, which

can be viewed as an exit strategy for exiting the CIP, should be based on projected

inspection workload and must provide for the necessary turnover of issues.

* -It is necessary to ensure that each phase of the preoperational inspection program is

properly completed. To the maximum extent possible, all issues (such as licensee test

exceptions or construction deficiencies) must be closed out before the programs are

officially considered complete. Items that are carried over into the operating phase must be

extensively documented, and the closure requirements for the items must be identified.

Inspection Program Structure and Implementation

The program must be structured to guide inspectors to inspect needed items and to provide a

coherent and simple method for them to record necessary information.

* Onsite inspections should begin during site preparation before the COL or CP is issued. A

continuous onsite inspection staff should be established and maintained throughout

construction. To ensure' that the full range of construction -activities- is-covered by

appropriately qualified inspectors, and because of the phased natureof many of those

activities, the mix of expertise among the resident inspection staff should be rotated.

* ' Inspection requirements should be made as objective as possible, allowing clear

determinations that critical attributes either have or have not been met. Establishing

B-2
discrete, objective inspection requirements will limit the need for subjective interpretations

of acceptability, and the sizable body of accumulated objective information will support major

inspection program conclusions.

* Objective inspection requirements should be established, to the maximum possible extent,

for systems, structures, and components, as well as for plant programs. Each inspection

procedure should clearly state how much inspection should be performed in order to

consider the procedure complete.

* Constructing a plant in a short period of time means that activities will happen rapidly and

in parallel with each other, which will place significant demands on inspection resources.

Planning and scheduling therefore need to be closely coordinated with plant construction

plans.

Inspection Documentation

At the end of the construction process, NRC must possess a fully documented body of inspection

data to support the findings that need to be made to allow plant operation.

* In some past construction projects, inspection reports did not fully document all areas that

had been evaluated during plant construction. The resulting incomplete inspection

documentation resulted in a lack of audit trails that could be used to respond to questions

raised during the process leading up to issuance of an operating license. Also, inspection

reports did not always clearly identify the items that had been inspected in the plant. The

revised CIP requires those individual samples (such as identification numbers for welds,

pipe supports, and cable terminations) be recorded in the CIPIMS. In addition, each

construction inspection in the future should be considered satisfactorily completed only after

supervisory or management personnel determine that the inspection is fully documented.

* In the past, NRC inspection reports focused generally on the deficiencies identified during

the inspections, without providing much detail on positive inspection findings. As a result of

such unbalanced inspection reporting, the NRC staff sometimes had to perform extensive

reviews during the final stages of plant licensing to provide additional information to support

licensing decisions. In some cases, the inspections had already been done but had not

been fully documented. To avoid follow-up reviews, future construction inspections should

document both satisfactory and unsatisfactory findings.

B-3
Appendix C

Inspection Sampling

ITAAC Sample Selection

A cornerstone of the Part 52 process is the concept of ITAAC which if met are necessary and

sufficient to provide reasonable assurance that the facility has been constructed and will be

operated in conformity with the license, the provisions of the Atomic Energy IAct, and the

Commission's rules and regulations. The staff will rely on the licensee to ensure that all of the

ITAAC have been met and will perform audit-type inspections to verify compliance with the ITAAC.

In performing these audit-type inspections the staff needs to address the fundamental question of

how much inspection is necessary to ensure that the acceptance criteria contained in the ITAAC

have been met.

Both the licensee and the NRC benefit from the process used during the COL and design

certification reviews that determine the level of detail for the ITAAC. For the designs that have been

certified (i.e., ABWR, System 80+, and the AP600) ITAAC were developed for the SSCs within the

scope of the designs. These ITAAC are part of the Tier 1 material found in the design control

document for these designs.

A process was developed to determine the level of detail for each ITAAC. This process is

discussed in sections 14.3 of the respective design control documents and also in the staffs draft

standard review plan section 14.3. The Tier 1 information has an entry for every system that is

either fully or partially within the scope the design certification. The intent of this comprehensive

listing is to define at the Tier 1 level the full scope of the certified design. However, the amount of

information in the Tier 1 entry, including the ITAAC, is commensurate with the significance of the

system. Several factors were used to determine the significance of the system including the

following:

* whether the feature or function is necessary to satisfy the NRC's regulations in 10 CFR

Parts 20, 50, 52, 73, or 100

* whether the feature or function pertains to a safety-related structure, system or component

* whether the feature or function represents an important assumption or insight from the

probabilistic risk assessment

whether the feature or function is important in preventing or mitigating a severe accident

whether the feature or function has had a significant impact on the safety or operation of

existing nuclear power plants,

whether the feature or function is typically the subject of a provision in the technical

specifications

. whether the feature or function in question is specified in the standard review plan as being

necessary to perform a safety-significant function

For many non-safety systems with low risk significance, the Tier 1 entry is limited to the systems'

name only. For this group, it is sufficient to ensure that the system has been completed before fuel

loading is allowed.

C-1
The staff believes that the process that has been developed and implemented for the certified

designs provides a good starting point for answering the fundamental question of how much

inspection is enough. That is, by having a construction inspection program that is ITAAC-focused

for the hardware portion of the design, the staff has already narrowed the field of inspection

activities. Operational programs, which do not have ITAAC associated with them are another matter

and are discussed in other sections of this document.

The staff does not intend to review or inspect every inspection, test, or analysis listed in every

ITAAC. To establish an NRC inspection footprint, the staff will ensure that, at a'minimum, it- has

received an ITAAC determination letter from the licensee for all ITAAC. If no inspections have been

performed related to that ITAAC as documented in CIPIMS (e.g., inspection findings or assessment)

the staff will review the' licensee's records for the ITAAC determination basis as necessary to

provide confidence that the ITAAC have been met.

Statistical Methods

This approach involves the development and implementation of statistical sampling methods with

the goal of obtaining, at the end of a plant's construction phase, a confidence statement about the

quality of plant construction. The October 1996, draft revised CIP report noted that the major

difficulty with applying statistical sampling to a nuclear power plant construction inspection program

would arise from the attempt to make confidence statements about the many non-homogeneous

processes that occur at a construction site.

The draft revised CIP report also referenced a memorandum to the Commission from E. Volgenau,

Director, Office of Inspection and Enforcementdated February 11, 1977, titled, Inspection Program

Utilizing Statistical Sampling Inspection Techniques." This memorandum discussed the results of

a'series of statistically based operating phase inspections that were performed at Three Mile Island

Unit 1. This trial program showed that strictly statistically based sampling was, on balance, not an

optimal method of inspection planning for three reasons:

* the statistical method identified no significant safety concerns that the traditional method

failed to identify;

* the traditional method successfully identified significant safety concerns that the statistical

method did not identify,

* and; the statistically based method was comparatively more resource-intensive.

However, the memorandum did note that confidence statements for a wide range of populations and

sample sizes could be developed for possible application to discrete portions of the inspection

program.

Since the time of the 1977 memorandum the staff has applied the use of statistical sampling

techniques to inspection-related activities. For example, resolution of some issues associated with

construction inspection of welding programs, and the dedication of commercial-grade items for use

in nuclear power plants relied on the use of statistical sampling techniques.

Regarding welding programs, shortly before the Seabrook full-power license was to be issued, the

NRC received a series of allegations, questions, and concerns about safety at the plant. Some of

C-2
the issues related to the adequacy of pipe welds made on-site during construction. The NRC used

statistical based sampling techniques to aid in its investigatiois and inspections of this issue. An

October 4, 1991, letter from James Taylor to the Commission titled, "Completion of the NRC Staff

Review of the Quality of ASME Field Welds at Seabrook," references these techniques.

Sampling techniques are also discussed in Draft Regulatory Guide 1070, uSampling Plans Used for

Dedicating Simple Metallic Commercial Grade Items for Use in Nuclear Power Plants." The

commercial-grade dedication in this draft regulatory guide refers to an acceptance process

undertaken to provide reasonable assurance that a commercial-grade item to be used as a basic

component in a nuclear power plant will perform its intended safety function and is deemed

equivalentto an item designed and manufactured under a quality assurance program in accordance

with 10 CFR 50 Appendix B. Although the draft regulatory guide is intended to provide guidance

for the development of a licensee's commercial-grade dedication programs, the staff believes that

some of the concepts developed for this program are applicable to the NRC's own construction

inspection verification programs.

An area where statistical sampling techniques could possibly be used for the construction inspection

program is welding. -Both the ABWR and the AP600 contain ITAAC associated with welding.

Statistical-based sampling techniques could be used for the staff to make findings and ITAAC

determinations for this process. If future plants were to have welding performed off-site (because

of modular construction techniques) and on-site, separate welding assessments could be made for

the off-site facilities as well as for on-site welding activities.

The staff could use statistical sampling techniques such that it will have a high confidence of a low

defect rate. To satisfy this premise, both the resolution of issues associated with the Seabrook

-:welding issue and DG-1070, used statistical sampling plans that result in at least 95 percent

confidence that populations with more than 5 percent defective items will be rejected. Because of

modular construction techniques it may be necessary to make assessments about several facilities

taken together. If this is the case then the samples at each facility can be adjusted to support the

staffs evaluation.

The staff believes that such statistically based sampling techniques are limited to certain areas. As

the revision to the construction inspection program moves forward, the staff hopes to identify ITAAC

and construction activities that lend themselves to such a technique and develop inspection

procedures that will provide guidance for how such techniques should be used.

Risk Informing Construction Inspection

The 1996 report on the revised CIP identified that PRA information could be used by the NRC to

perform sensitivity, uncertainty, and importance analyses to identify those -plant SSCs whose

passive failure (due to inadequate construction) would most greatly impact the plant's risk profile.

In this way, the more risk-significant SSCs would be identified, and construction inspection samples

could be skewed toward those SSCs. -

The selection of ITAAC were heavily risk informed during the design certification process because

design-specific PRA is required as part of a design certification in accordance with 10 CFR

C-3
r

52.47(a)(v). These PRAs were used during the applicant's development, and the staffs review, of

the ITAAC for the designs that were certified.

The Use of Risk in Developing ITAAC for AP600

The AP600 was chosen as an example because the safety systems for this design use passive

means (such as gravity, natural circulation, condensation and evaporation, and stored energy) for

accident prevention and mitigation. These passive safety systems perform safety injection, residual

heat removal and containment cooling functions. In this design, traditional active systems like the

emergency diesel generators are non-safety related.

Section 14.3 of the AP600 design control document provides background information on the

selection criteria for how the ITAAC were developed. The selection criteria consisted of deterministic

and PRA based inputs. Table 14.3-1 of the AP600 design control document provides the results

of the ITAAC screening summary. The screening of the 90 AP600 systems led to several systems

not being selected for an ITAAC. In addition, for the AP600 there are 32 systems, such as the

potable water system, turbine building closed cooling water system, and the heater drain system,

were only the system is listed in the Tier 1 material. The end result is that 39 of the original 90

systems (greater than 40%) that were screened resulted in no detail ITAAC being developed.

While many systems were screened out for consideration during the ITAAC development several

systems were included and different aspects of those systems augmented in the ITAAC because

of risk insights. As mentioned earlier the emergency diesel generator is non-safety related for the

AP600 design, however, there are ITAAC associated with the EDG because of it's risk significance.

Similarly, there are non-safety related functions of the normal residual heat removal system that

have ITAAC associated with them, in part because of their risk significance. Table 14.3-6 of the

AP600 design control document contains the design features from the PRA perspective that were

considered important to verify in ITAAC. Because of the information in the design certification the

staff has a good starting point for the use of risk in the inspection program for the designs that have

been certified.

The staff intends to use the design-specific PRA's to help further focus its ITAAC inspection

activities. While such risk information will be useful in developing construction inspection samples

and focusing on audit activities, the actual conduct of construction inspections will primarily

represent a deterministic process. This is important because a plant must be built in accordance

with its design criteria for subsequent PRA usage to be valid .

The Role of the Quality Assurance Program

Quality assurance (QA) and quality control (QC) will be an integral part of the NRC's inspection

effort, and will be a common component of the inspections that are performed by the staff, in that

10 CFR 50 Appendix B applies to construction activities done in accordance with 10 CFR Part 52.

The staff believes that one of the major lessons learned from past nuclear power plant construction

efforts is that the identification of construction problems, and the timeliness and extent to which they

are corrected are effective measures of licensee management's control over onsite activities.

C4
NRC experience shows that, if the licensee deals thoroughly with corrective action, including the

identification and correction of root causes, there is a good 'hance that the overall quality of the

construction is good. If these areas are weak, it is likely that there are lapses in quality; such a case

would be evident if repetitive problems occur.

The role of quality assurance was emphasized in SECY-00-0092, 'Combined License Review

Process," dated April 20, 2000. The following is paraphrased from this SECY paper:

The NRC staff anticipates that there will be design, construction, and testing

activities related to ITAAC verification for which the staff will not be able to rely solely

on NRC inspections to verify proper completion. For these activities, the staff must

rely on the licensee's QA program to provide suitable controls for effective

verification. The staff must have confidence that the licensee's QA program is

adequate and that it is being properly implemented so that design, construction, or

testing deficiencies are identified, documented, and corrected. The QA

requirements of Appendix B to Part 50 apply to all safety-related activities being

conducted by the licensee during the design, construction, and operations phase,

including those safety-related activities performed to satisfy ITAAC. For example,

preoperational test program testing performed to demonstrate that safety-related

structures, systems, and components (SSCs) will perform satisfactorily in service

must be conducted under a program that satisfies Criterion Xl, "Test Control," of

Appendix B. It may also satisfy testing required by the ITAAC process. The scope

of the initial test program, however, is not limited to just safety-related SSCs.

Specifically,-Regulatory Guide (RG) 1.68, Initial Test Programs for Water-Cooled

Nuclear Power Plants," specifies the scope of plant SSCs to be tested to satisfy the

-requirements of Criterion 1, "Quality standards and records," of Appendix A, and

Appendix B to Part 50. Although testing is required for all SSCs within the scope of

RG 1.68, it is not required that all of them be tested to the same stringent

requirements. Accordingly, the administrative requirements that govern the conduct

of the test program contain provisions for the application of administrative controls

in a manner commensurate with the safety significance of the SSCs within its scope.

Because the ITAAC process includes safety-related activities that must be

conducted under a QA program that meets the requirements of Appendix B to Part

50, licensees must develop programmatic controls and procedures that delineate

how such activities will be implemented.

As discussed in public meetings with NEI representatives, there may be deficiencies

identified by the QA program that are relevant to ITAAC and that must be addressed

by the licensee before-the NRC can find that the ITAAC have been successfully

completed. NEI 'representatives asserted that QA and QC deficiencies have no

relevance to ITAAC findings. The'NRC staff disagrees with any assertion that

QA/QC deficiencies have no relevance to the determination of whether ITAAC have

been successfully completed. Simply confirming that ITAAC had been performed

in some manner and a result obtained apparently showing that the acceptance

criteria had been met would not be sufficient to support a determination that ITAAC

had been successfully completed. The manner in which ITAAC are performed can

be relevant and material to the results of the ITAAC. For example, in conducting

C-5
ITAAC to verify a safety-related pump's flow rate, it is necessary, even if not

explicitly specified in the ITAAC, that the gauge orinstrument used to verify the

pump flow rate be calibrated in accordance with the requirements of Appendix B to

Part 50 and that the test configuration be representative of the final as-built plant

conditions (i.e., valve or system lineups, gauge locations, system pressures, or

temperatures). - Otherwise, the acceptance criteria for. pump flow rate could

apparently be met while the actual flow rate in the system could be different than that

required by the approved design. Therefore, the NRC staff has determined that a

QA/QC deficiency may be considered in determining whether an ITAAC has been

successfully completed if (1) the QA/QC deficiency is directly and materially related

to one or more aspects of the relevant ITAAC (or supporting Tier 2 information) and

(2) the deficiency (considered by itself, with other, deficiencies, or with other

information known to the NRC) leads the NRC to question whether there, is a

reasonable basis for concluding that the relevant aspect of the ITAAC has been

successfully completed. This approach is consistentwith the NRC's current methods

for verifying initial test programs.

The NRC staff recognizes that there may be programmatic QA/QC deficiencies that

are not relevant to one or more aspects of a given ITAAC under review and,

therefore, should not be relevant to or considered in the NRC's determination as to

whether that ITAAC has been successfully completed. Similarly; individual QA/QC

deficiencies unrelated to an aspect of the ITAAC in question would not form the

basis for an NRC determination that an ITAAC has not been met. Using the ITAAC

for pump flow rate example, a specific QA deficiency in the calibration of pump

gauges would not preclude an NRC determination of successful ITAAC completion

if the licensee could demonstrate that the original deficiency was properly corrected

(e.g;, analysis, scope of effect, root cause determination, and corrective actions, as

appropriate) or that the deficiency could not have materially affected the test in

question. Furthermore, during the development of ITMC,- the design certification

applicants determined that it was impossible (or extremely burdensome) to provide

all details relevant to verifying all aspects of ITAAC (e.g., QA/QC) in Tier 1 or Tier

2. Therefore, the NRC staff accepted the applicants' proposal that top-level design

information be stated in the ITAAC to ensure that it was verified, with an emphasis

on verification of the design and construction details in the "as-built" facility. To

argue that consideration of underlying information, which is relevant and material to

determining whether ITAAC have been successfully completed, is not necessary

ignores this history of ITAAC development.

In the September 5, 2002, staff requirements memorandum associated with SECY-00-0092, the

Commission approved the staff's recommendation that underlying information (such as QANQC

deficiencies), which is relevant and material to ITAAC,; must be considered in determining whether

ITAAC have been successfully completed. In addition, there may also be deficiencies identified that

are not relevant to ITAAC. These deficiencies may still need to be addressed by the licensee, but

they will not necessarily delay a finding on successful ITAAC completion or plant operation.

In summary, the staff believes that statistical sampling and PRA techniques can be used as an aid

to help to focus its ITAAC-based inspection efforts. The staff also believes that inspections of

C-6
QA/QC (especially corrective action program inspections) will be an important aspect of the review.

This paper presents a high-level approach for using these methods. Details of the design need to

be known to employ these techniques properly.

The staff plans to take selected examples from the designs that have been certified and develop

these techniques further. The results of these examples will be made publicly available and will be

used as aids in the development of the detailed inspection procedures for a COL. The staff intends

to delay work on revising the detailed inspection procedures until it has more information from the

industry on the details and the design for any particular nuclear power plant that may be constructed

in the future.

C-7
Appendix D

Examples of Information That Would

Invalidate a Previous ITAAC Determination

This appendix gives examples for discussion of what the NRC staff considers what constitutes!new

and -significant information." The examples show how "new and significant information" might

impact a previously accepted ITAAC.

Example 1: A test instrumentation QA/QC deficiency directly related to whether an ITAAC

acceptance criterion had been met.

The role of quality assurance was emphasized in SECY-00-0092, "Combined License Review

Process," dated April 20, 2000.-The following is paraphrased from this SECY paper.

The manner in which ITAAC is performed can be relevant to its results. For example, in verifying

an ITAAC associated with a safety-related pump's flow rate, the gauges or instruments used to

verify the pump flow rate must be calibrated in accordance with the requirements of 10 CFR Part

50 Appendix B. The test configuration must also be representative of the final as-built plant

conditions. For example, valve or system lineups, gauge locations, system pressures, and

temperatures must be in accordance with the design. Otherwise, the acceptance criterion for a

pump flow rate could apparently be met while the actual flow rate in the system was different from

the required design flow rate. The NRC staff therefore determined that a QAIQC deficiency may

be considered in determining whether an ITAAC has been successfully completed.

A QA/QC deficiency could be relevant if the deficiency is directly related to the ITAAC or its

supporting Tier 2 information. The deficiency by itself or with other deficiencies, may lead the NRC

to question whetherthere is a reasonable basis forconcluding that the relevant aspectofthe ITAAC

has been successfully completed. This approach is consistent with the NRC's current methods for

verifying initial test programs.

Example 2: Improper weld materials used in the fabrication of an ITAAC related structure. (from

Appendix E)

The example involves the reactor pressure vessel support platform for Taipower's Lungmen-1

ABWR. Improper welding material was used for initially assembling the platform. The 1,000-ton

platform support was made of steel-reinforced concrete and the steel portion was manufactured at

the China Shipbuilding Corp. (CSC) in Kaohsiung in southern Taiwan. The platform was shipped

to the site and then assembled by CSC personnel at the Lungmen site.

-Workers initially used low-strength welding material to assemble the platform, instead of highstrength material specified by the engineering codes. The welding material was confirmed to be

inappropriate and Taipower will have to reassemble the platform.

-The information concerning the use of improper welding material could have come in the form of

an allegation to the licensee, or to the NRC. In such a case, the NRC would investigate and if the

allegation were substantiated the NRC would evaluate the impact on any associated ITAAC.

D-1
Appendix E of this document discusses this example in more detail including the ITAAC that may

be impacted.

Example 3: IE Bulletin No. 83-07, Apparently Fraudulent Products Sold by Ray Miller, Inc., and

NRC Bulletin No. 88-05, "Nonconforming Materials'Supplied by Piping Supplies, Inc. (PSI) at

Folsom, New Jersey and West'Jersey Manufacturing Company (WJM) at Williamstown, New

Jersey."

Bulletin No. 83-07 was issued after the NRC completed a review of records that were in the custody

of the U.S. Attorney's office, and determined that materials with fraudulent documentation had been

supplied to nuclear power plants.

Bulletin No. 88-05 was issued after the NRC obtained copies'of certified material test reports

(CMTRs) for material supplied by PSI and WJM that contain false information about material

supplied to the nuclear industry. 'A donestic forging company's letterhead was apparently used on

a number of CMTRs to certify that commercial-grade and foreign steel met the requirements of

ASME Code Section III, Subarticle NCA-3800. There was no evidence that PSI orWJM performed

or had a subcontractor perform' the testing required by Section III to upgrade the commercially

produced steel for these falsified CMTRs.

Information of this nature concerning the construction materials for ITAAC-related components or

structures could be considered significant enough to invalidate a previous ITAAC determination.

Example 4: NRC Bulletin No. 92-01, "Failure of Thermo-lag 330 Fire Barrier System to Maintain

Cabling in Wide Cable Trays and Small Conduits Free From Fire Damage"-

This bulletin notified licensees of failures in fire endurance testing associated with the Thermo-Lag

330 fire barrier system installed to protect safe shutdown capability,-and requested all operating

reactor licensees to take recommended actions.

During construction, information of this nature concerning fire protection materials for safety-related

or risk-significant systems could be considered significant enough to invalidate a previous ITAAC

determination.

Example 5: NRC Bulletin No. 88-10, "Nonconforming Molded-Case Circuit Breakers"

NRC Information Notice (IN) 88-46, "Licensee Report of Defective Refurbished Circuit Breakers,"

and Supplement 1 thereto, reported that Anti-Theft Systems,: Inc., a local electrical'distributor,

supplied 30 circuit breakers (CBs) to the Diablo Canyon nuclear power plant. These circuit

breakers (Square D molded-case, type KHL 36125) were'intended for use in non-safety-related

applications at Diablo Canyon. Square D Company' reported that inspection and testing of these

CBs determined that they were refurbished Square D Company equipment.- Furthermore, Square

D reported that several of the circuit breakers tested did not comply with Square D or Underwriters

Laboratories, Inc. (UL) specifications for all of the electrical tests performed.- IN 8846 also listed

several California companies' that were involved in supplying surplus, and possibly defective,

refurbished electrical equipment to the nuclear industry.

D-2
During the NRC inspections of defective refurbished circuit breakers, additional examples were

identified that indicate a potential safety concern regarding eiectrical equipment supplied to nuclear

power plants. The NRC was concerned that equipment being procured as new, and assumed to

meet all applicable plant design requirements and/or original manufacturer's specifications may not

conform to these requirements and specifications.

While the bulletin discussed CBs supplied for non-safety-related applications, it is now understood

that non-safety-related electrical systems can be risk-significant. Information of this nature,

concerning potentially defective circuit breakers for safety-related or risk-significant systems, could

be considered significant enough to invalidate a previous ITAAC determination.

D-3
Appendix E

ABWR Construction Example

This appendix provides an example of a construction activity that could be done off site that could

impact inspections, tests, analyses and acceptance criteria (ITAAC). The example involves the

Lungmen-1 advanced boiling water (ABWR) reactor that is being constructed in Taiwan. The

design is very similar to the design that was certified by the NRC and that is codified in Appendix

A of 10 CFR Part 52. However, because of licensing differences, there are no ITAAC associated

with the Lungmen design. The staff believes that if the problem that was encountered in Taiwan

happened in the United States, it would directly impact an ITAAC.

Description of the Problem

The problem involves the reactor pressure vessel support platform for Taipower's Lungmen-1

ABWR. Improper welding material was used for initially assembling the platform. The 1,000-ton

platform holding the vessel is made of steel-reinforced concrete and was manufactured at the China

Shipbuilding Corp. (CSC) in Kaohsiung in southern Taiwan. The platform was shipped to the site

and then assembled by CSC personnel at the Lungmen site.

Workers initially used low-strength welding material to assemble the platform instead of

high-strength material specified by engineering codes. The welding material was confirmed to be

inappropriate and Taipower will have to reassemble the platform.

The vessel support platform has five layers of reinforced concrete. The initial problem manifested

itself when a hairline crack about 50 centimeters (cm) long and between 0.2 cm and 0.3 cm deep

was discovered in the lowest level of the platform. The platform weighs 464 metric tons, is 13

meters (m) high and has a diameterof 14 m. At the time of the discovery Taipowerfiled a so-called

quality assurance "noncompliance" report with the regulator. It was subsequently determined

through an inspection that improper welding material was used for the assembly. The inspection

was performed after irregularities in welding were suspected on the site.

U.S. Design Control Document Information

The design control document for the U.S. ABWR is incorporated by reference into Appendix A of

10 CFR Part 52. The design control document consists of Tier 1 and Tier 2 information and generic

technical specifications. The Tier 1 material consists of the following:

* definitions and general provisions

* design descriptions

* ITAAC

* significant site parameters

* significant interface requirements

E-1
The following information is extracted from Section 2.14.1, "Primary Containment System," of the

U.S. ABWR Tier 1 information:

The RPV pedestal forms the lower drywell region and consists of a cylindrical double

shell composite steel structure. It is anchored to the basemat and'supports the RPV

through a support ring girder. The pedestal also supports the reactor shield wall. The

pedestal consists of two concentric steel cylinders joined together radially by vertical

steel diaphragms and filled with concrete. The pressure suppression venting paths

are an integral part of the pedestal structure, which includes (1) the ducts which

interconnect the lower and upper drywell regions, (2) the vertical downcomers from

the interconnecting ducts to the horizontal vents, and (3) the horizontal vents that

direct steam into the suppression pool. The horizontal vents consist of 30 pipes

uniformly spaced around the perimeter of the pedestal in ten stacks of three each.

The total horizontal vent area is greater or equal to 11.55 m2.The distance from the

pedestal containing these horizontal vents to the outer suppression pool wall is

greater than 7.4m. All HVAC ducts, cabling and piping between the upper and lower

drywells are routed through the interconnecting ducts

The ITAAC that could be affected (if the problem occurred in the U.S.) are 2.14.1.3 and 2.14.1.14

(shown below).

Table 2.14.1 Primary Containment System

Design Commitment Inspections, Tests, Analyses Acceptance Criteria

3. The ASME Code pressure 3. A structural integrity test. 3. The results of the SIT of

boundary components of the (SIT) will be conducted on the pressure boundary

PCS will retain their integrity, the pressure boundary components conform with the

under internal pressures that components of the PCS per requirements of the ASME

will be experienced during ASME Code requirements. Code.

service.

14. The containment internal 14. A structural analysis will 14. A structural analysis

structures are able to be performed which report exists which concludes

withstand the structural reconciles the as-built data that the as-built internal

design basis loads as defined with structural design as structures are able to

in Section 2.14.1. defined in Section 2.14.1. withstand the design basis

loads as defined in Section

2.14.1.

E-2
A diagram of the ABWR primary containment system is also contained in the Tier 1 material as

follows (note 2 in the figure refers to the location of the reactor pressure pedestal):

Lessons Learned for the Construction Inspection Program

The staff believes that this example shows the need to perform offsite inspections to support a

determination by the staff that an ITAAC has been completed. As described in the main body of the

framework document, the staff expects to perform inspections of the facility that fabricates the

reactor pressure vessel pedestal. The inspections associated with this offsite fabrication facility

would include but not be limited to the following:

- a quality assurance inspection of the licensee to ensure that the details of the contract

properly reflect the importance of this structure and clearly identify the quality control

requirements for the structure

- inspection of the offsite facility to ensure that the structure being properly manufactured,

including a review of welding records and quality assurance and quality control

- inspection of the onsite assembly of the modular structure, including a review of the

processes to ensure that the structure was not damaged during shipping, assembly by the

contractor, and final placement.

E-3
Appendix F

Generic Combined License From SECY-00-0092

[NAME OF NUCLEAR FACILITY]

[NAME OF NUCLEAR FACILITY OWNER]

Docket No. 52-[XXX]

License No. NPF-[XX]

The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for a combined license (COL) filed by [name of nuclear facility

owner(s) (the licensee)][, which references Appendix _ to 10 CFR Part 52,]

complies with the standards and requirements of the Atomic Energy Act of 1954, as

amended (the Act), and the applicable regulations set forth in 10 CFR Chapter I,

and all required notifications to other agencies or bodies have been duly made;

B. The applicable requirements set forth in 10 CFR 52.77, 52.78, 52.79, 52.81, 52.83,

52.85, 52.87, 52.89, [52.91, if applicable], and 52.97 [and Appendix _to 10 CFR

Part 52] have been met;

C. There is reasonable assurance that the facility will be constructed and will operate

in conformity with the application, as amended, the provisions of the Act, and the

applicable regulations set forth in 10 CFR Chapter I, except as exempted from

compliance in Section 2.F below;

D. There is reasonable assurance that the activities authorized by this COL can be

conducted without endangering the health and safety of the public and (ii) that such

activities will be conducted in compliance with the applicable regulations set forth in

10 CFR Chapter I, except as exempted from compliance in Section 2.F below;

E. The licensee is technically and financially qualified to engage in the activities

authorized by this COL in accordance with the applicable regulations set forth in

10 CFR Chapter I;

F. The licensee has satisfied the applicable provisions of 10 CFR Part 140, "Financial

Protection Requirements and Indemnity Agreements."

G. The issuance of this license will not be inimical to the common defense and security

or to the health and safety of the public;

H. The issuance of this license is in accordance with 10 CFR Part 51 and all applicable

requirements have been satisfied; and

F-1
I. The receipt, possession, and use of source, byproduct, and special nuclear material

as authorized by this license will be in accordance with the applicable regulations in

10 CFR Parts 30, 40, and 70.

2. On the basis of the foregoing findings regarding this facility, COL No. NPF-[XX] is hereby

issued to [licensee], to read as follows:

A. This license applies to the [Name of Nuclear Facility], a light-water nuclear reactor

and associated equipment (the facility), owned by the licensee. The facility is

located and is described in the licensee's final safety analysis report (FSAR), as

supplemented and amended, and the licensee's environmental report, as

supplemented and amended.

B. Subject to the conditions and requirements incorporated herein, the Commission

hereby licenses the licensee:

(I) Pursuant to Sections 103 and 185.b of the Act and 10 CFR Part 52, to

construct, 'possess, use, and operate the facility at the designated location

in accordance with the procedures and limitations set forth in this license;

(2) Pursuant to the Act and 10 CFR Part 70, to receive and possess at any time,

special nuclear material as reactor fuel, in accordance with the limitations for

storage and amounts required for reactor operation, described in the FSAR,

as supplemented and amended;

(ii) Pursuant to the Act and 10 CFR Part 70, to use special nuclear material

as reactor fuel, after the finding in Section 2.D(1)-of this license has been

made, in accordance with the limitations for storage and amounts required

for reactor operation, and described in the FSAR, as supplemented and

amended;

(3) Pursuant to the Act and 10 CFR Parts 30, 40,' and 70, to receive, possess,

and use, at any time, any byproduct, source, and special nuclear material as

sealed neutron sources for reactor startup, sealed sources for reactor

instrumentation and radiation monitoring equipment calibration, and as

fission detectors in amounts as required;

(4) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess,

and use in amounts as required, any byproduct, source, or special nuclear

material without restriction to chemical or physical form, for sample analysis

or instrument calibration or associated with radioactive apparatus or

components; and

(5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not

separate, such byproduct and special nuclear materials as may be produced

by the operation of the facility.

F-2
C. The license is subject to, and the licensee shall comply with, all applicable provisions

of the Act, and the rules, regulations, and orders of the Commission, including the

COL inspections, tests, analyses, and acceptance criteria (ITAAC) contained in

Appendix A of this license.

D. The license is subject to, and the licensee shall comply with the conditions set forth

in 10 CFR Chapter I, now or hereafter applicable [consistent with the requirements

in Section VIII of Appendix _to 10 CFR Part 52]; and the conditions specified and

incorporated below:

(I) Nuclear Fuel Loading

The licensee shall state under oath or affirmation to the Commission that the

acceptance criteria in the COL ITAAC have been met.

(ii) The licensee is authorized to load fuel into the reactor vessel and perform

precritical testing (zero power) after the Commission has found, in

accordance with 10 CFR 52.103(g), that the acceptance criteria have been

met.

(2) Low-Power Testing

Upon approval of the Director of the Office of Nuclear Reactor Regulation,

the licensee is authorized to perform low-power testing and operate the

facility at reactor steady-state core power levels, not in excess of [XX]

megawatts thermal (5-percent power), in accordance with the conditions

specified herein.

(3) Maximum Power Level

Upon approval of the Director of the Office of Nuclear Reactor Regulation,

the licensee is authorized to perform power ascension testing and operate

the facility at reactor steady-state core power levels, not in excess of [XXXX]

megawatts thermal (100 percent power), in accordance with the conditions

specified herein.

(4) Incorporation

The COL ITAAC, plant-specific Technical Specifications, Environmental

Protection Plan, and Antitrust Conditions contained in Appendices A, B, C,

and D, respectively, of this license are hereby incorporated into this license.

E. The licensee shall report any violations of the requirements in Section 2.D of this

license within 24 hours. Initial notification shall be made in accordance with the

provisions of 10 CFR 50.72, with written follow up in accordance with the procedures

described in 10 CFR 50.73.

F-3
F. The following exemptions are authorized by law and will not endanger life or

property or the common defense and security. Certain special circumstances are

present and these exemptions are otherwise in the public interest. Therefore, these

exemptions are hereby granted.

[(1) LISTING OF EXEMPTIONS FROM DESIGN CERTIFICATION RULE (DCR)]

[(2) LISTING OF EXEMPTIONS WHICH ARE OUTSIDE THE SCOPE OF DCR]

G. The licensee shall fully implement and maintain in effect all provisions of the physical

security, guard training and qualification, safeguards contingency plans, and all

amendments made pursuant to the authority of 10 CFR 50.90, 50.54(p), 52.97[, and

Section VIII of Appendix -to Part 52] when nuclear fuel is first received onsite, and

continuing until all nuclear fuel is permanently removed from the site.

H. The licensee shall have and maintain financial protection of such type and in such

amounts as the Commission shall require in accordance with Section 170 of the

Atomic Energy Act of 1954, as amended, to cover public liability claims.

I. The following operational requirements that are applicable to this license will become

effective after the Commission finds that the acceptance criteria in this license (COL

ITAAC) have been met in accordance with 10 CFR 52.103(g):

(1) emergency plans,

(2) technical specifications,

(3) .......

J. After the Commission has made the finding required by 10 CFR 52.103(g), the COL

ITAAC [not including the Tier 1 information from the referenced design certification

rule (DCR)] do not 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

proceeding.

K. This license is effective as of the date of issuance and shall expire at midnight on

[the date 40 years from the date of issuance];

FOR THE NUCLEAR REGULATORY COMMISSION

Director

Office of Nuclear Reactor Regulation

Appendices:

Appendix A - COL ITAAC (including Tier 1 information]

Appendix B - Technical Specifications [plant-specific]

Appendix C - Environmental Protection Plan

Appendix D - Antitrust Conditions

Date of Issuance:

F-4
Appendix G

An example of an NRC Organization for Implementing

a Construction Inspection Program

The CIP team has proposed the following example organization to the CIP steering committee.

The basic organization is similar to the Special Projects organizations of the late 1980s and

early 1990s for TVA and Comanche Peak.

The CIP implementation may be performed using different organizations to address onsite

construction and remote fabrication of systems, structures, and components. All activities could

be coordinated through the onsite organization and be planned, scheduled, and tracked through

completion using the CIP Information Management System (CIPIMS).

The basic construction inspection organization would exist on site. This organization would -

consist of six individuals, supported by regional and headquarters technical experts and

inspectors. The organization structure is presented in Figure G-1 below.

The onsite organization would be lead by the site construction inspection supervisor (SCIS).

,This senior staff member (GG-15) would be responsible for all onsite NRC personnel and any

associated activities involving NRC or NRC contract support personnel. The SCIS would report

to the director of the division of reactor projects. The SCIS would serve as the staff inspection

supervisor and senior resident inspector for much of the construction period, until the region

determined that the site inspection activities and senior resident responsibilities needed to be

separated. This individual would serve as the senior NRC staff member on site regardless of

visitors or temporary assignees. All direct communication between the applicant and the

regions or headquarters would be required to go through the SCIS. If a senior staff member for

the applicant wanted to speak directly with the region, the individual would notify the SCIS prior

to contacting the region. The SCIS would be directly responsible for all inspection activities

performed on site.

The onsite inspection team would consist of a scheduler and three team leaders, one each for

-mechanical inspection activities, electrical and instrument and control (EIC) inspection activities,

and civil and structural inspection activities. Miscellaneous inspection activities will be

distributed amongst the team leaders by the SCIS. (See figure below.)

The scheduler would be trained in a scheduling program that is compatible with the applicant's

scheduling software. This individual would also be trained on the use of the CIPIMS and would

be responsible for the overall utilization and maintenance of the CIPIMS for that site. All efforts

should be made to assign this individual for the duration of the project. The scheduler would be

directly responsible for interfacing with the applicant's scheduling organization and coordinating

all inspection activities with the onsite inspection team leaders. The scheduler would report

directly to the SCIS. -

The three team leaders would be responsible for all onsite inspection activities relating to his/her

assigned discipline(s). Team leaders would be fully qualified with sufficient experience to

G-1
perform and/or supervise the inspections within their purview. They would be-responsible for

ensuring that all resources are available and scheduled prior to scheduled inspection activities.

They would ensure that technical and/or inspection support personnel are properly briefed and

appropriately prepared for the planned inspection prior to the start of any inspection activities.

Team leaders must be trained on the CIPIMS and would be ultimately responsible for

developing and processing inspection reports, ensuring that the CIPIMS is properly updated,

and preparing Federal Register notifications for onsite inspections, as appropriate. For any

extenided-absences (in excess of one week), team leader' coverage should come from outside

the site organization. In addition to the onsite construction inspection team, a resident inspector

woud be as~signed early on in the construction process. This individual would assume the

routine resident responsibilities anfd interactions with the applicant' 'and would -serve'as a

replacement for team members during routine absences and limited annual leave. This

individual wo uld be trained on the use of the CIPIMS, and become sufficiently familiar with

scheduling activities to be able to replace the team scheduler for short periods. The resident

inspector would be directly assigned to the SCIS until the region determined the need to assign

a separate senior resident inspector.

Remote inspections or fabrication inspection/audit activities would be implemented by different

implementing organizations that would coordinate their activities with the onsite construction

inspection team using the CIPIMS. All remote inspection/audit activities would be coordinated

by the director of the Division of Inspection Program Management. In general, inspection/audit

activities for major U.S. manufacturers' would be assigned to the region responsible for thegeographical area Where the manufacturing is occurring. Foreign manufacturer

inspections/audits generally would be performed out of headquarters. Special

inspectionis/audits generally would be performed out of headquarters or "centers of excellence.'

Centers of excellence are organizations, such as the regions, NRR, RES, and possibly NMSS,

other Govemnment agencies, or specialized consultants that have a concentration of specialized

skills to perform inspections/audits of unique production activities. These production activities

may' involve software, monitoring instrumentation, fuel fabrication, safeguard components, etc.

If multiple specialized skills are needed for a single inspection/audit, then a mix of organizations

may be assigned to a single remote inspection/audit. If so, consideration needs to be given to

the nature of the inspection/audit,' and the organization that is providing the most support to the

effort in determining the lead organization. The lead organization would be determined by

senior executives representing each of the organizations involved.; Difficulties in deciding the

lead organization would be resolved by the Director, Division of Inspection- Program

Management. The lead organization would assign a team leader. The team leader would be

responsible for the development of the audit/inspection report and updating the CIPIMS through

the onsite organization. If headquarters were assigned as the lead organization, the appropriate

project manager from the Division of Licensing Project Management would update the CIPIMS

for the team. However, the ultimate responsibility for updating the CIPIMS would remain with

the team leader. All remote audit/inspection reports will require SCIS concurrence.

G-2
All onsite and remote Inspection/audit information would be placed in the CIPIMS within 45

working days of completing the inspection. Updating of CIPIMS information would be completed

within 15 working days of issuing the applicable inspection/audit report.

Nuclear Regulatory Commission Construction

Inspection Program Organization Chart

G-3
Appendix H

Comparison of AP600 and ABWR ITAAC to Inspection Procedures

SUMMARY

One objective of the Construction Inspection Program Team was to determine the scope of

effort required to develop the procedures that will be necessary to verify that the ITAAC have

been met. To meet this objective, the team obtained contract support from three experienced

engineers. One had significant experience in construction and operational inspections, another

had operational inspection experience, and the third had significant design experience.

The team tasked the contractors to review the ITAAC for the AP600 and ABWR designs and

compare the acceptance criteria to the existing NRC inspection procedures. The purpose of this

review and comparison was to identify any glaring holes that would require'significant effort to

develop procedures for guidance-on inspecting and verifying the completion of ITAAC.

The result of the effort was encouraging. There were relatively few ITAAC that would require

new procedures. Many of the existing procedures would require some revision to be fully

capable of verifying the ITAAC, while some procedures were acceptable as written.

This appendix describes the contractors' methods'and results.' This provides a starting point for

the development of the procedures that will be necessary to support the construction of new

nuclear power plants under the regulations of 10 CFR Part 52. A future review will be performed

to develop matrices for the two approved designs to identify all procedures that would be

necessary for the verification of each ITAAC.

1.0 INTRODUCTION

Beckman & Associates, Inc. was contracted to evaluate the adequacy of NRC inspection

procedures (IPs) in implementing the Construction Inspection Program (CIP) for new facilities.

The CIP for facilities to be constructed in accordance with approved design control documents

was evaluated by comparing the acceptance criteria listed in each design control document to

the guidance available in IPs. The design control documents for the Westinghouse AP600

design and the General Electric ABWR design were reviewed.

The design control documents contained Tier 1 specifications for the systems, facilities, and

programs that had been approved by the NRC. The Tier 1 documents included tables

associated with the required Inspections, tests, analyses, and acceptance criteria (ITAAC). The

acceptance criteria will require inspection verification as part of the CIP.

The three Beckman team members developed a spreadsheet for each of the two plant types

that were evaluated. The spreadsheets summarize each of the acceptance criteria listed in the

design control document, the IP determined to be applicable to inspect that acceptance criterion,

and an evaluation of the adequacy of the guidance contained in the IP to fully inspect the

acceptance criterion.

H-1
The Beckman team reviewed and evaluated IPs selected from all of the NRC inspection

programs (Construction, Pre-Operational Testing and Operations Phases) that appeared to

contain guidance that could be used to verify the adequacy of an acceptance criterion.' Most of

the IPs provided to the Beckman team members were from the NRC CIP team leader. Other

procedures were obtained from the NRC electronic database, and some were found in personal

hard copy files.

Some procedures listed in an older, (November 1993) NRC' Inspection Manual Index had titles

suggIesting that the procedu res-might contain applicable guidance, but copies of the lPs could

not be located. Those~ I Ps are mIenitioned under the applicable spreadsheet heading but were

not credited for providing guidance.

The guidance provided in the IP selected for a given acceptance criterion was compared to the

required verification activity to establish which areas had adequate inspection guidance and

where additional guidancelwas needed. The adequacy- of the IP to verify the acceptability of the

acceptance criterion was then categorized using the following criteria:

Category 1: -There was essentially no IP that addressed the Acceptance Criterion.

Category 2: ,The IP required major revision to fully address the Acceptance Criterion.

Category 3: The IP required minor revisions to fully address the Acceptance Criterion.

Category 4: The IP was essentially adequate to evaluate the Acceptance Criterion.

In assigning the category designation-, Beckman team assumed that the construction verification

activities would be completed during normal construction inspection activities. The construction

verification would, in some cases, also fulfill the verification of completion of the acceptance

criteria. The team tended to select a higher category designation for some IPs that did not

provide detailed guidance for an attribute when the associated acceptance criteria required a

specific verification activity.

Some of the same lPs were frequently found to contain inspection guidance, applicable to more

than one of the ITAAC areas. The category designation that was assigned to the IP in one area,

or for one acceptance criterion within an area, was based on the particular requirement.

2.0 SUMMARY OF FINDINGS

The selected category designations listed in the spreadsheets were tabulated for each of the

reviewed plants.' When more than one category. had been proposed by team members, the

caegr wsdetermined on a weighted basis considering the requirement an te guidne

The number of acceptance criteria listed in each of the categories is shown below:

H-2
DESIGN CATEGORY

1 2 3 4

AP600 136 215 478 9

ABWR 99 115 975 233

These numbers show that 351 (42% of the total) acceptance criteria for the AP600 design and

214 (15% of the total) acceptance criteria for the ABWR design lack significant inspection

guidance.

The details of the above findings were further evaluated for each design.

2.1 AP600 Summary

Sections 2 and 3 of the AP600 Tier 1 design control document presented the descriptions and

ITAAC tables for the approved systems, facilities, and programs. There were 93 areas in

Sections 2 and 3, but no ITAAC were associated with 37 of the areas. A total of 838 acceptance

criteria was listed in the ITAAC tables for the remaining 56 areas.

No applicable [Ps were discovered for three of the AP600 ITAAC Tables: 2.3.19,

"Communications"; 3.2, "Human Factors Engineering"; and 3.7, "Reliability Assurance Program."

Those sections of the design control document were also listed at the top of the spreadsheet.

All of the other tables containing acceptance criteria were found to have at least partial IP

guidance available for some of the listed requirements.

While few areas were considered to have complete, standalone guidance available, the IPs

designated for over half of the total number of AP600 acceptance criteria were considered to be

Category 3. For many of the items classified as Category 3, incorporating instructions in the

designated IP to review the design control document and providing guidance to verify the

acceptance criteria, could produce complete guidance.

There were 351 AP600 acceptance criteria that were determined to have little or no available IP

guidance and were, therefore, judged to be Category 1 or Category 2. Although an applicable

IP was designated for over half of these acceptance criteria (the 215 Category 2), the guidance

provided in the available IP was considered to be lacking in detail or specificity and would

require significant revisions and/or additions.

It should be noted that the number of Category 2 entries does not directly represent the number

of IPs that require revision. The same IP was frequently listed as the applicable guidance for a

number of different ITAAC requirements. The following IPs were listed as Category 2 guidance

for numerous acceptance criteria:

H-3
IP TITLE # of Acceptance

Criteria

37051 Verification of As-Builts 9

50100 Heating, Ventilating, and Air Conditioning 7

Systems

70434 Engineered Safety Features Actuation 22

System Test- Preoperational Test

Witnessing

70444 Containment Isolation Valve Test - 9

Preoperational Test Witnessing

93807 Systems Based Instrumentation and Control 17

Inspection

The number of IPs that would require major revision to meet the Category 4 criteria is, therefore,

considerably less than the total number of Category 2 acceptance criteria. There was not -

enough time to analyze IPs that were identified number of IPs in each category. It appears,

however, that the number of Category 2 IPs requiring significant revision would' be less than 35.

No applicable inspection guidance was located for 136 AP 600 acceptance criteria (16%). A.

review of the details for these category 1 acceptance criteria determined that many of the

deficient areas could be grouped together and addressed in generic IPs. A listing of the generic

IPs follows:

* Procedure for basic configuration/functional arrangement

* Procedure for'review of certified stress reports

* Procedure for a hydrostatic/pressure test

* Procedure for mechanical separation of divisional equipment

* Procedure for motor-operated valves, check valves, power-operated valves, valve failure on

los's of motive power

* Procedure for review of seismic analysis to withstand safe shutdown earthquake

* Procedure for review of structural analysis to withstand design basis accident loads

* Procedure for review of maintenance of containment integrity/isolation

* Procedure for review of ASME Design Report

* Procedure for review of ASME non-destructive examination (NDE) Report

* Procedure for review of Leak-Before-Break Report

* Equipment qualification pr6cedure (mechanical and electrical)

* Procedure for physical separation and electrical isolation of test signals

* Procedure for test signal and displays, parameters, and controls in main control room

H4
As examples of areas where similar requirements could be grouped together, 35 acceptance

criteria required the review of ASME, NDE, and hydrostatic test results were designated

Category 1 because specific IPs were not located. These areas could probably be inspected

using generic IPs that provide guidance for that type of inspection. Likewise, a generic

procedure for evaluating the environmental qualification of mechanical equipment could be

developed to address the 11 mechanical environmental qualification acceptance criteria that

were determined to be Category 1 because of the lack of an appropriate IP. Numerous other

acceptance criteria required verification that a report was available to show the design value(s)

had been met.

In addition, a final review of the AP600 spreadsheet found that 24 acceptance criteria required

verification of specific details (e.g., tank volume, flow rate, or orifice size) that would require little

additional inspection guidance. Specific items could be covered by a 'generic' ITAAC

inspection procedure that provided guidance on reviewing the design control document and

verifying the acceptance criteria listed in tables.

Therefore, it would appear that minor changes to existing procedures and the development of a

small number of generic IPs would decrease the number of AP600 Category I items from 136 to

approximately 50.

2.2 ABWR Summary

The ABWR design control document was reviewed using the same methodology that had been

used to review the AP600 document. The total number of ABWR tables and associated

acceptance criteria was much larger than those provided for the AP600 design. There were 159

tables in Sections 2 and 3 of the ABWR design control document with a total of 1422

acceptance criteria listed. (No acceptance criteria were listed in 62 tables)

No applicable IPs were discovered for eight of the ITAAC tables in the ABWR design control

document. Those ITAAC are listed below. All of the other ITAAC tables that listed acceptance

criteria were found to have some amount of IP guidance available.

2.3.2 Area Radiation Monitoring Systems

2.111 1 Station Service Air System

2.11.12 Instrument Air System

2.11.13 High Pressure Nitrogen Gas Supply System

2.12.16 Communications

2.14.6 Atmospheric Control System (N2 Injection)

H-5
3.1 Human Factors Engineering

3.6 Design Reliability Assurance Program

The tabulation of the categorization of inspection guidance for the ABWR acceptance criteria is

provided above. A large percentage (69%) of the items were considered to be Category 3 and,

like the AP600 Category 3 items, many of those could be upgraded to Category 4 with minor

revisions to existing procedures.

There were 214 ABWR acceptance criteria that were listed as Category I or Category 2-15%

of the total number. While this number indicates that many new IPs are needed and major:

revisions to others are required, the number was much lower than for the AP600. I The smaller

number of items that the team determined had only Category 1 or 2 guidance available was

considered to be mainly attributable to the number of ABWR acceptance criteria that involved

specific verification actions (e.g., verify valve opens on signal, or verify opening time is less than

3 seconds). The team's familiarity with the IPs may have also led to the somewhat higher

ratings of the procedures.

The ABWR spreadsheet was also found to contain a number of IPs that had been determined to

provide only Category 2 guidance for a number of different ITAAC requirements. The following

IPs are examples:

IP TITLE # of inspection

criteria

70432 Control Rod System Test - Preoperational 17

Test Witnessing

93807 Systems Based Instrumentation and Control- 15

Inspection

In addition, the same IPs that were designated Category 2 for some attributes were noted to be

rated as providing Category 3 or 4 guidance for others. The number of IPs requiring major

revision to provide the Category 4 guidance for ABWR acceptance criteria was, therefore, also

considered to be less than the total number of Category 2 attributes.

The ABWR acceptance criteria were reviewed again to determine areas where similar

requirements could be grouped together so that inspections could be conducted using general

guidance. The findings were similar to the AP600 findings. For instance, there were nine

requirements to verify specific values that could be inspected using a generic ITAAC verification

IP. There were also 20 acceptance criteria for system hydrostatic/pressure test results

verification that could be inspected using one or two general leak test IPs.

H-6
Based on an overview of the.ABWR acceptance criteria designated as Category 1 and Category

2, it appears that the number! of IPs that need to be develope'd and/or significantly revised is

around 60.

3.0 CONCLUSIONS

There were 351 AP600 and 214 ABWR acceptance criteria identified as having insufficient

inspection guidance provided by existing IPs. Those acceptance criteria were designated as

Category 1 and Category 2 in the applicable spreadsheet. As discussed above, however, the

number of acceptance criteria lacking sufficient inspection guidance does not directly

correspond to the number of IP changes that would be needed. The same IP was frequently

found to provide varying levels of guidance for different acceptance criteria. Therefore, the

revision of one IP could improve the rating of a number of Category 1, 2 and 3 entries.

In addition, many of the same IPs; were found to be applicable to both plant designs that were

reviewed. Therefore, one revision could affect the category rating of a given IP in more than

one location, in more than one plant design. Constraints did not allow a determination of the

number of these "common' IPs.

The total number of new IPs that need to be developed and existing IPs that require major

revision was considered to be around 100.

Most of the existing W~s will require at least slight revisions to include guidance for reviewing the

applicable section(s) of the design control document and to verify completion of any related

acceptance criteria.

An evaluation of the total number of IPs that were identified in the spreadsheets and an analysis

of the necessary additions would help to determine the total level of effort required to produce

acceptable levels of inspection guidance for all of the acceptance criteria.

H-7
Appendix I

Comment Resolution Summary for the

Draft Construction Inspection Program Framework Document

NOTE: Comments followed by an item number refer to items in the letter and its attachments

(ML 033090096) provided by NEI and endorsed by a letter from Southern Company

(ML033090101).

Comments followed by a page number are comments/questions from the transcript of

the August 25 2003 public workshop on the framework document which the NRC

committed to treated as public comments. (ML032790347)

Comments followed by (AC) refer to recommendations from the Advisory Committee

on Reactor Safeguards (ML033510735)

Comment I Recommendation/ Resolution

Question

Topic: IMC-2501 - ESP

-1 How does the NRC intend to ensure Contractors employed by the NRC for either

that contractors used by the NRC are headquarter's staff review of an ESP application

trained to the new standards etc? or participation in a Region lead inspection will

(Page 47) receive orientation training before beginning

work. Additionally, such contractors will receive

on-going direction by -the NRC ESP project

manager or Regional inspection team leader.

However, no change was made to the document

because this is at a greater level of detail than

intended for the Framework Document.

2 Will Part 21 be applied to an applicant?

(Page 42)

3 In IP 35002, for Part 21, what is the

;,.applicability to contractors?,(Page 38)

4 Reconsider the applicability of Part 21

to ESP applicants. (Letter item 1)

New, Research and Test Reactors (RNRP) staff

and the Office of the General Counsel (OGC)

are aware of the issues surrounding the

applicability of Part 21 to ESP applicants and are

reviewing the matter.

5 Modify IP 35002 to eliminate -the

reference to Part 21 applicability to

ESP applicants. (Item A1.2)

1-1
Comment I Recommendation/ Resolution

Question -_X

6 Revise paragraph B2 on page 7 to The paragraph has been revised to remove the

reflect that an ESP provides approval misleading statement.

of a site for one or more plants and

may not expire when a COL or CP is

issued. (Item A1.3)

7 What are the expectations regarding The Emergency Preparedness and Plant

Appendix B? (Page 32) Support Branch issued a letter to NEI on August

4, 2003 which stated that the NRC expects that

an ESP, applicant -will use a QA control

framework equivalent in substance to that

described in Appendix B. Review Standard (RS)

8 Is a deviation or deficiency in -002 provides one method for demonstrating that

accordance with Appendix B something the ESP QA controls are equivalent in substance

that the applicant would have to to Appendix B. However, an applicant's failure to

address?(Page 51) use a framework equivalent to Appendix B would

not, in and of itself,: result in rejection of the

application. Further, any deficiencies or

9 Will the NRC staff request a QA deviations in an applicant's quality assurance

program through other means? (Page measures would need to be addressed to ensure

54) the reliability and integrity of data contained in or

supporting the ESP application. The staff

believes that Section 17.1.1 of the review

standard allows sufficient flexibility for an

10 Modify the framework document to applicanttoproposealternateQAmeasures, and

clarify the expectation that ESP QA that no change to the text is necessary as a

measures be equivalent in substance result of these comments.

to Appendix B. Clarify also in IP 35002

and Review Standard-002. (Item Al. 1)

Topic: Construction Inspection Program Information Management System

(CIPIMS

11 How are the licensees going to share

construction schedules if the

information contains proprietary

details? (Page 92)

Meeting(s) will be scheduled in the future to

explore further how information could be shared.

12 How and how often is the NRC going

to get information from the licensee and

vendors to support CIPIMS? (Page

1 09)

1-2
Comment I Recornrmendation/ Resolution

Question

13 Schedule follow-up interaction to

discuss coordination of construction

and inspection schedules, use "of

coding schema, protection of

proprietary information and business

sensitive schedules. (Letter item 2 and

Items A3.1.1, A3.1.2,A3.1.3)

Tpic: IMC-2502 - r-OL Phase T.

14: Clearly identifythat a principal objective No change is needed since the EDV inspection

of NRC EDV is to provide reasonable objective stated in the framework document

assurance that detailed design quotes SECY-94-294. Verifying that an

information on which construction will applicant's design is consistent with the

be based is consistent with the design approved design, during either design

approved during a design certification certification or COL review, is one purpose of the

or COL review. (Item A2.1.1) EDV.

15 Modify the framework document to SECY-94-294 indicated that -as plant

reflect that the engineering design construction progresses, NRC will determine if

verification (EDV) need not :be the engineering design is adequate. The NRC

completed prior to COL issuance. (Item will use performance-based inspections to verify

A 2.1.2) that plant systems and components are installed

and tested to applicable standards, certified

design information, and ITAACs. Thus, the EDV

inspections would continue during construction

after the COL application had been approved.

The wording in the framework document was

revised to reflect that the staff plans to conduct

these inspections as early in the process as

practical but that they may continue after the

COL is issued.

116 Modify the framework document to

reflect that the scope of EDV may

encompass review of additional topical

design areas such as fire protection,

environmental qualification, seismic

design, HELB analyses, and

separation/independence. (Item

A2.1.3.1)

No change made. The framework document will

not provide an all inclusive list of the different

inspection areas reviewed by the EDV

inspections. Design areas like those referenced

are unique to the applicant's specific design and

theirassociated ITAACs would denote the design

requirements to be met on a per system basis.

Topical reports submitted by the licensee would;

be reviewed during the COL application stage.

1-3
Comment I Recommendation/ Resolution

Question

17 Include public notification via a Federal No change made. The NRC will use inspection

Register notice or other method of the reports as the primary vehicle to inform the

NRC determination that the licensee licensee and the public about the result of

design engineering processes are inspection efforts, including the results of the

acceptable. (Item 2.1.3.2) NRC's review of the design engineering process.

18 Include information stating that afterthe The information about the design engineering

NRC has approved the design process has been revised to reflect that the NRC

engineering process, follow-up does: not approve the, applicant's design

inspection to spot check the process engineering process, but rather assesses its

would: focus on configuration viability. The NRC. inspections will determine

management and design details whether the applicant abides by that process in

completed after the main thrust of NRC actuality. The NRC will adjust its inspection effort

engineering 'design verification 'was either up or down based on its determination of

completed and had established the whether or not the design engineering process is

acceptability of the licensee's overall working as expected.

design engineering processes. (Item

A2. 1.3.4)

19 Incorporate the idea that the main

thrust of' NRC engineering design

verification would focus on design

areas other than those covered by DAC

(e.g., piping, instrumentation and

control, 'and the main control' room),

unless the applicant chose to complete

and seek NRC approval in the COL of

all or a portion of the plant design in

such DAC areas. After staff reviews in

areas with DAC are complete, EDV in

these areas may be accomplished as a

follow-up to the main EDV milestone

achieved at the time of COL issuance

or early in construction. Or, perhaps

more likely, the staff safety reviews and

EDV may occur in parallel as the plant

design in DAC areas is completed.

(Item A 2.1.3.4)

No change made. Areas currently covered by

DAC, such as those described in SECY-94-294,

where design descriptions and functional system

drawings are adequate for licensing reviews but

not for actual construction or construction

inspection activities, would be appropriate for

inspection. The timing of the reviews of DAC will

depend on when the applicant actually provides

sufficient information to warrant the inspections.

1-4
Comment / Recomrnibndation/ . -. Resolution

Question

20 Clarify that it is at least possible and Document revised to reflect that rather than

perfectly acceptable under Part 52 that waiting until the end of construction, the NRC will

a COL applicant might not contract for start its review as soon as it can based on the

major components, detailed design applicant's readiness.

engineering or construction until after a

COL is issued. (Item A2.2)

21 Distinguish between licensing reviews

of operational program descriptions

based on the SRP or other COL

application review guidance versus

operational program 'readiness

inspections prior to plant operation in

accordance with IMC-2504.' This

-distinction should be 'made in the

Framework document regardless of the

outcome of parallel-'interactions

concerning the extent of operational

program information to be provided in

COL applications. Regardless of the

outcome of those interactions, the

Framework document should reflect the

focus of IMC-2504 on inspections to

determine operational' program

readiness prior to operation. (Item

A2.3)

The document has been revised to identify that

the operational programs will receive two

evaluations. The first will be a review of the

program description as part of the COL

application. The second evaluation will be an

inspection that will take place prior to plant

operation and will focus on the licensee's

readiness to implement the program.

1-5
Comment / Recommendation/ Resolution

Question

Topic: IMC-2503 - ITAAC Verification

22 If an inspection of ITAAC activities The section related to ITAAC determinations has

indicatesthattherearedeficienciesthat been revised to. incorporate this

have not been addressed by the recommendation.

licensee's corrective action program,

then the licensee should be required to

identify and correct the weakness in the

correction action program that lead to

the deficiency. (AC)

23 SAYGO ITAAC -- What does that The use of a SAYGO approach was designed to

mean? (Page 140) help. the staff support the Commission in their

What does 'sign' mean in sign as you determination that the ITAAC have been met.

go? (Page 156) Because the. existing ITAAC are based on

complete systems being installed and do not

recognize the multiple stages of construction, the

use of SAYGO is envisioned as a tool by which

the staff can signify overall satisfaction with the

licensee activities that have been completed on

an ITAAC to that point leading up to the

24 The staffis creating ITAC where there completion of an ITMC. The staff has selected

is not really an ITAAC (Page 151) that approach because it will allow for an on- going assessment to determine if the staff is on

track, to complete the necessary and sufficient

reviews to support the 10 CFR Part 52.103(g)

finding by the Commission.

The description of SAYGO in the framework was

revised to incorporate the key points of this

explanation.

25 What would a 52.99 notice look like?

(Page 159)

No change made. The Framework document, by

design, will not contain detailed information,

rather it acknowledges what will exist within the

overall program. Specific information on form

and format of NRC documentation will be

covered in implementing procedures including

manual chapters and inspection procedures.

1-6
Comment I Recommendation/ Resolution

Question

26 How would the staff go about making The document was revised to reflect that the

the recommendation to the office recommendation to the NRR office director or

director or regional administrator?, regional administrator will be based on a review

Would it be mechanical or is it of the information associated with ITAAC

gathering additional information to includingassuringthatadetermination letterwas

supplement or a whole new received, reviewed, accepted, and has been

assessment? (Page 184) noticed in the Federal Register by the NRC.

27 Modify the framework document to Language revised to eliminate the use of

reflect that SAYGO process 'conclusion'. The section -now shows that

conclusions provide confidence in the programs and processes are part of the

acceptability of quality-related evaluation.

construction processes, including

conformance with applicable codes and

standards, QA Program requirements,

etc. (Item A 3.2.1)

28 Expand the list of example processes No change made. The table on page 17 is not

identified on p. 17 of the framework intended to be all inclusive.

document to include still other

construction-related processes that

may be amenable to early, systematic

assessment and determination of

acceptability by the NRC, such as

receipt inspection, commercial grade

dedication, warehousing and others.

(Item A 3.2.1)

29 Either clarify that SAYGO ITAAC See comments to 23 and 24.

conclusions are SAYGO process

conclusions that correspond directly to

ITAAC acceptance criteria or the

concept of SAYGO ITAAC conclusions

could be eliminated. (Letter item 5 and

Item A3.2.2) .

30 The distinction between ITAAC

conclusions by the NRC staff and the

Commission's ITAAC finding is clear

without the word Ointerim." Consider

using the term 'Section 52.99 ITAAC

conclusions,' (Letter item 3 and Item

A3.2.3)

Change made. The term "ITAAC conclusion"

has been changed to 'ITAAC determination."

1-7
Comment / Recommendation/ Resolution

Question

31 Expand the ' discussion regarding The document has been revised to indicate that

independent review of -ITAAC the NRC intends to complete a review of the

verifications to include clarifications that documentation presented by the licensee in the

the independent review of the ITAAC ITAAC determination letter as it is received. This

completion would not involve re-review review will ensure that the NRC has considered

of all' ITAAC but rather-would be a every ITAAC. The staff will also review any

'vertical slice audit' and could begin in inspection information related to each specific

advance of fuel load and in parallel with ITAAC when it is submitted by the licensee as

ITAAC 'verification activities in' the complete.

region. (Item A3.3.1)

When the licensee has informed the NRC that

they have completed all of the ITAAC, a final

independent review will consist of an audit that

will ensure that the' NRC' has -received a

determination letter for each ITAAC, agrees that

it is complete, and has published the required

Federal Register notification in accordance with

52.99.

32 Define and incorporate the process for The information ;was incorporated into the

triggering the Section 52.103(a) notice document. The added information reflects the

consistent with the description from the NRC's November 20, 2003 response to NEI on

November 2001 NEI white paper. their November 2001 white paper. (ADAMS

(Page 189, Letter item 6 and Item A Accession No. ML032760053)

3.3.2)

33 Include a statement that NRC ITAAC

documentation, including Section 52.99

notices, should focus on the licensee's

ITAAC determination bases. Matters

not material to ITAAC determinations

would be the subject of normal NRC

inspections and reports. (Letter item 6

and Item A3.3.3)

The document has been revised to indicate that

the "normal NRC inspections" are the means by

which NRC will establish confidence in the

licensee's construction program. In addition to a

completeness review of the documentation

submitted by the licensee with the ITAAC

determination, inspection results' related to that

ITAAC will 'establish the bases for NRC

acceptance of the licensee's ITAAC

determination bases.

I-8
- t - ,.-G

Comment I Recommendation/ Resolution

Question

34 Revise the informfation related to The licensee's corrective action program will be

rescinding a prior ITAAC conclusion to an integral part of addressing inspection issues.

reflect an approach that would rely on The framework document has been revised to

the licensee's corrective action reflect that the NRC envisions that in most cases,

program to address most issues items identified by the NRC will be turned over to

affecting installed system, structures or the licensee to be addressed through the

components that arise after an ITAAC corrective action program. However, if NRC

is complete and a 52.99 notice is identifies new and significant information that

issued. (Item A 3.3.4) calls a previous ITAAC determination into

question, we will consider rescinding it. This

decision would not be taken lightly, and would be

a deliberate NRC management decision.

Although the licensee may use their corrective

action program processes to address the reason

for rescinding any ITAAC determination, the

nature of this decision will call for the NRC to

closely monitor how the licensee resolves the

issue.

35 The staff should reserve public Agree, no revision to the document needed.

meetings to exchange information

regarding ITAAC deficiencies for

situations when there are particularly

significant negative findings

necessitating involvement of NRC and

licensee senior management. (Item A

3.3.5)

Topic: - Non-hTAAC IMC-2504 Insprections

36 The Framework document should be

modified to reflect that IMC-2504 will

begin after the COL is issued. IMC2504 should focus on (1) non-ITAAC

inspections prior to fuel load (primarily

ORAT inspections) that will support

Region and NRR recommendations

regarding readiness to load fuel, and

(2) post-fuel load inspections prior to

power operations (primarily start-up

testing inspections). (Item A4.1)

The title of IMC-2504 has been revised from

"Preparation for Operations" to "NON-ITAAC

Inspections" to better reflect the the range of

inspection activities to be covered.

The scope of inspections to be conducted under

the new title will include the inspection activities

to begin before the COL is issued.

1-9
Comment / Recommendation/

Question

Resolution

37 The gap between the. ITAAC

completion and the point at which tech

spec surveillance requirements

become effective should not be

referred to as the regulatory gap since

the licensee would. be implementing

QA, design control, work control,

configuration control during that time.

(Page 200)

Document revised to remove 'regulatory gap"

38 On p. 24 of the Framework document,

the staff uses the term regulatory "gap"

to describe the time between when an

individual ITAAC is complete and when

the Commission makes it's Section

52.103(g) finding and, discusses the

need for inspections to ensure that the

licensee is "managing this 'gap'

appropriately. While such inspections

may be appropriate, it is incorrect and

misleading to refer to a regulatoryugap,"

and we recommended that the staff use

different terminology to describe these

inspections. (Item A 4.2)

1-10
Comment I Recommendation/ | I . Resolution

Question

Topic: Specific Comments on the Draft Framework Document

39 On page 5 clarify (1) that ITAAC can Document was revised to clarify how SAYGO will

be shown to be complete only after the be used.

underlying construction, inspection and

test activities are complete. This

necessarily means that demonstration

of ITAAC completion will occur later in

construction for some ITAAC versus

others. And (2) ITAAC verification by

the NRC will be based on SAYGO and

other NRC inspection conclusions that

are material to the ITAAC conclusion.

(Item B. 1)

40 In the discussion of IMC-2502, clearly Because design acceptance criteria are ITAAC,

state that the control room ITAAC and the only requirement is that they be completed

other "design acceptance criteria" are prior to fuel load. If designs are not complete

not required to be completed at time of before the COL is issued, then they will be

COL issuance. (Item B.2) carefully reviewed as they are completed. A

provisions for design inspections will be included

in IMC-2504.

Document revised to indicate that these ITAAC

could be completed prior to COL.

41 Correct reference on page 11 in section No change made. Reference is correct as

D.1 to read Section 52.79(b)(1). (Item written in the framework document.

B.3)

42 If ITAAC information is used as an Agreed

example the acceptance criteria should

be stated verbatim. (Item B.4.)

43 Modify the statement on page 19 to

read as follows: 'Upon receipt of an

ITAAC determination letter, the NRC

staff will base its decision regarding

ITAAC acceptability on a review of the

licensee's ITAAC determination record

I and/or-en NRC inspection reports and

NRC SAYGO documentation that are

material to the ITAAC in question.'

(Item B.5.)

[Changes have been made to the statement to

reflect the intent of this comment.

I-11
Comment / Recommendation/ Resolution

Question

44 On p. 19, the Framework Document The document has beenlrevised to reflect at a

says 'the staff will perform' an minimum the NRC''will review the available

independent review to ensure that it information to -ensure that' the agency has

has received an ITAAC determination received, reviewed, accepted, and published a

letter for each ITAAC and the staff notice in the Federal Register foreach and every

agrees that all the ITAAC have been ITAAC.

met." This language should be

modified to reflect the purpose as In addition, a sample of the ITAAC packages

clarified by the NRC staff during the may be selected for review as a further

August 27 workshop to audit and assurance of the accuracy of the data.

independently verify the ITAAC

verification activities of the primary

regional inspectors. As discussed in

comment A.3.3.1, above, a 100% reverification is not envisioned; it is

expected that sampling and vertical

slice audit methods would be used by

the independent review team. (Item

B.6.) , ,

45 On page 23 of the Framework

Document, we recommend this

statement be modified as follows

pending the final resolution of the

programmatic ITAAC issue: 'Therefmre,

if Regardless of whether or not an

operational program does not-have has

an ITAAC, there is an expectation that

the staff will perform inspectionspriorto

operation to verify the licensee's

compliance with regulations."

(Item B.7.)

Document revised to incorporate the intent of the

comment but not the exact wording stated.

1-12
Comment I Recommendation! Resolution

Question

46 By definition, 'operational program Document reviseed to incorporate the intent of the

inspections under IMC-2504 are comment but not the exact wording stated.

separate from ITAAC ' verifications

under IMC-2503. Therefore, on page

23 of the Framework Document, we

recommend this statement be modified

as follows: "To the extent these

[transition to ROP] inspections are

performed prior to loading fuel, these

inspections will also supplement the

bases for the regional administrator's

recommendation to the Director of NRR

regarding IRTAAGlant readiness to

load fuel." This change is consistent

with language on p. 24 regarding

consideration of ORAT results. (Item

B.8.)

47 'On page 24, the Framework Document

'-states that programs such as technical

'specifications must be in place and fully

functional priorto the 52.103(g) finding.

This statement is not accurate and

conflicts with an earlier statement on

the 'same page, which states that

- technical specifications will not become

effective until the' NRC issues its

52.103(g) finding. This page should be

revised to indicate that the licensee

must be ready to implement the

technical specifications and other

applicable operational programs prior

to the 52.103(g),' and not that they be

"fully functional"' before that finding.

(Item B.9.)

No change was-made because the document

states that - uPrior to the Commission findings,

the staff expects the licensee to, phase in

programs such as technical specification controls

so that problems are recognized and solved

before the program is required (by regulations or

license condition) to be fully implemented."

1-13
Comment I Recommendation/ Resolution

Question

48 On p. 25 of the Framework Document, The document was modified to reflect the staffs

the staff envisions separate NRC current understanding that the COL will contain

authorizations after the Commission license conditions. The document reflects how

makes its 52.103(g) finding to go above recommendations will be made if conditions are

5% power and to full power. The staff part of the license.

notes that the Commission approved -

these authorizations in the SRM on The underlying issue of whether or not a COL

SECY-00-0092. However, the staff should include conditions and what the form of

recommendation and the Commission any such conditions :should be will not be

approval of separate low- and full- resolved in the framework and has been turned

power authorizations occurred without over to the New Reactor Licensing Section.

discussion with stakeholders of

whether these actions are consistent

with Part 52 and before the impact of

these actions could be fully

explored. (Item B.10.)

49 A target, such as 30-days from receipt No change made. Assigning a target for

of an ITAAC determination letter from completion is premature. However, the staff

the licensee, should be established for should establish due dates for completion of the

NRC to complete the ITAAC verification work as it is received for review. Assigning due

process and issue the required 52.99 dates will ensure that the work is timely but also

notice. (Item B.1 1) considers the, overall volume of received items .

The actual process, to be used for reviewing,

accepting and noticing an ITAAC determination

package will be detailed in an NRC implementing

procedure such as a manual chapter.

Topic: Other comments and questions i

50 Will the Inspection Manual Chapters b

issued for public comment? (Page 21!

re The Inspection Manual Chapters will not be

5) issued for public comment. The Inspection

Manual Chapters 'will reflect the approach

outlined in the final Construction Inspection

Program Framework Document, on which the

public was provided the opportunity to comment.

1-14
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER

(2-89) (Assigned by NRC, Add Vol., Supp. Rev., NRCM 1102, and Addendum Numbers, If any.)

3201.3202 BIBLIOGRAPHIC DATA SHEET

(See instructions on the reverse)

2. TITLE AND SUBTITLE NUREG-1789

10 CFR Part 52 Construction Inspection Program Framework Document

3. DATE REPORT PUBLISHED

MONTH i YEAR

April 2004

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

7. PERIOD COVERED (Inclusive Dates)

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC. provide Dvision Offce or Region. U.S. NuclearRegulatory Commission. and mailing address;ff contractor,

provide name and mailing address.)

U.S. Nuclear Regulatory Commission

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (f NRC, type 'Same as above' if contractor, provide NRC Division, Office or Region, U.S. NuclearRegulatory Commission,

and mailing address.)

Office of Nuclear Reactor Regulation

U.S. Nuclear Regulatory Commission

Washing ton, DC 20555-001

10U. SUPIPILlMENIAY NUOITE

11. ABSTRACT (200 &voids orless)

The information contained in the Construction Inspection Program Framework Document, NUREG-1789, details the overall

philosophy and approach that will be used to inspect new nuclear power plants being licensed and built under 10 CFR Part 52.

The information contained in this NUREG will guide the development of Inspection Manual Chapters and Inspection Procedures that will be used to implement the construction inspection program.

12 KEY WORDSIDESCRIPTORS (s wios or phrases mart assist ear tessn AD -angthe reportj 13 AVAILABILITY STATEMENT

inspection unlimited

construction 14 SECURITY CLASSIFICATION

licensing (7ahs Page)

nuclear power plant unclassified

(T7hs Repot)

unclassified

15. NUMBER OF PAGES

16. PRICE

NRC FORM 335 (2-89) I his form was electronically produced by Elite Federal Forms, Inc.
NUREG-1789' 10 CFR PART 52 CONSTRUCTION INSPECTION PROGRAM APRIL 2004

FRAMEWORK DOCUMENT

UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

OFFICIAL BUSINESS