ML20078A461

Readiness Review Program Module 16 - Nuclear Steam Supply Sys

ML20078A461
Person / Time
Site:	Vogtle
Issue date:	12/03/1985
From:	Ramsey W GEORGIA POWER CO.
To:
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References
PROC-851203, NUDOCS 9406010203
Download: ML20078A461 (300)

Text

Georg a Power Company l Pro;ect Management l Route 2 Box 299A Waynesboro, Georgia 30830 fee phon (! 404 724 8114 404 554 9961 fV 1

L Vogtle Project i /~%

DU December 3, 1985 Mr. D. O. Foster Vice President and General Manager Vogtle Project Waynesboro, GA 30830 RE: Readiness Review Program Module 16 Nuclear Steam Supply System LOG: RR-606 FILE: X7BDlO2 I% h

Dear Mr. Foster:

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Pursuant to your instructions I am enclosing Module 16 of the l Readiness Review Program entitled Nuclear Steam Supply System. l This module reports the work of the Readiness Review Team and has been prepared in order to present you with an accurate picture of the readiness for operations of the Vogtle Project, based upon a close examination of the nuclear steam supply system.

The Readiness Review process included an initial assessment and review of basic licensing documents in order to identify Project commitments within the scope of the module. The Readiness

,s Review Team then verified implementation processes designed to meet those commitments, including programs and controls relating to work within the scope of the module.

The team then engaged in a process designed to verify that implementation programs were operating ac described in procedures and other descriptive documents. In concluding this verification process, the team then actually verified that the licensing commitments and the procedure and specification requirements identified were complied with.

9406010203 940512 PDR ADOCK 05000424 P PDR

Mr. D. O. Poster December 3, 1985 Page 2 q We are confident that the verification methodology used allowed the Readiness Review Team to properly appraise the actual Q condition of the nuclear steam supply system, and provided a valid means of assessing the quality of the program having also considered applicable past audits, inspection reports, and problems experienced by other utilities.

Based on the examinations, inspections, and evaluations of the review and the responses and corrective actions committed to by the Project, it is the conclusion of the Readiness Review Team that the design and construction programs that govern the nuclear steam supply system have produced a final product that meets design requirements and licensing commitments.

Additionally, none of the findings identified, either individually or collectively, are such that the adequacy of the project nuclear steam supply system is called into question.

Therefore, the nuclear steam supply system meets the FSAR commitments.

Members of the Readiness Review Team and I are prepared to O discuss this module with you at your convenience. If we can b provide you with any further information or assistance regarding this matter, contact me.

l Very truly yours, e .

William C. Ramsey .-- ,

WCR/deg l cc: R. E. Conway

'- Readiness Review Board Members Heading File Document Control 0079p/ 322- 5

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!.e VOGTLE ELECTRIC GENERATING PLANT i UNIT 1 4

4 j READINESS REVIEW 4,

i MODULE 16 - NUCLEAR STEAM SUPPLY SYSTEM S

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PREFACE r Georgia Power Company (GPC), in order to gain added assurance of the operational readiness of the Vogtle Electric Generating Plant (VEGP), is conducting a pilot Readiness Review Program.

The VEGP pilot Readiness Review Program is a systematic, in-depth self-assessment of work processes and verification of compliance with regulatory commitments. To accomplish the VEGP

() pilot Readiness Review Program, the work processes and regulatory commitments were divided into manageable segments called modules. There are approximately 20 modules. Each module is a predefined scope of VEGP activities.

4 Each module is intended to provide a brief description of the O method of complying with project licensing commitments pertaining to the module scope and is not intended to make further commitments or to revise in any way prior commitments.

If any differences exist between the commitments discussed in this document and the licensing documents, they are unintentional: and the licensing document governs.

Activities common to several modules are provided as General Appendixes. There are approximately 10 appendixes. These appendixes, as appropriate, are referenced in the modules and i are augmented in each module with module-scope-specific details as needed.

r The VEGP Readiness Reviaw Program is being conducted on a schedule to provide ad'.ed operational readiness assurance to GPC management in support of the VEGP Unit 1 operating license.

However, conclusions reached regarding programmatic and technical adequacy through review of VEGP Unit 1 are indicative of Unit 2, since both units are being designed and constructed together under a single quality assurance program: with like

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management controls, procedures, etc.; 'and.to the same specifications and criteria.

Stone and Webster Engineering Corporation has been contracted to provide technical management for, and technical personnel to implement, an independent design review as a part of the Readiness Review program. Additionally, Stone and Webster is O- reviewing project responses to Readiness Review findings for technical adequacy.

The VEGP Readiness Review Program is not intended to eliminate or to diminish any authorities or regulatory responsibilities now assigned to or exercised by the Nuclear Regulatory Commission or GPC.

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Further, the Readiness Review Program is not intended to change the techniques of inspections or assurance of quality program activities. Rather, the VEGP Readiness Review Program is an added program initiated by GPC management to assess the VEGP and to provide additional feedback to management O m

so that they may initiate any needed corrective actions in an orderly and timely manner.

The scope of work processes and regulatory commitment compliance covered by each module will be assessed by, and the module prepared and reviewed by, individuals collectively familiar with the design, construction, and operational processes of nuclear power plants. It is the collective opinion of the Readiness Review Task Force, Readiness Review Board, and GPC management that, based on their experience, the methodology used in the module process will assess, on a programmatic basis, the adequacy of project commitment implementation.

Readiness Review Discrepancy Reports and resulting dispositions are reviewed by the Readiness Review Program quality assurance staff and are input into the normal project process for safety significance and potential reportability evaluations in accordance with regulatory requirements.

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i EXECUTIVE

SUMMARY

l Introduction This module documents a review program to ascertain whether design and construction activities associated with Vogtle-specific aspects of the nuclear steam supply system (NSSS) comply with licensing commitments and whether compliance with the commitments is verifiable with existing project documentation. ,

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The scope of Module 16 for design includes an assessment of design control programs for the interfaces between the architect engineer (Bechtel Power Corporation (BPC)] and the NSSS supplier' l (Westinghouse). Technical evaluation of calculations and other design activities for'the NSSS interface will be included in a separate Independent Design Review report. For construction,-

this module addresses the installation of the primary loop components (reactor, steam generators, reactor coolant pump, pressurizer, etc.) by Nuclear Installation Services Company (NISCO) and' associated support activities (material control, storage, field construction engineering, etc.).

n The program consisted of two separate reviews: design program verification and construction program verification, O In implementing the above reviews, project documents such as design criteria, specifications, and procedures were reviewed along with results of past audits and inspections. In addition, the Readiness Review Board's technical consultant provided independent technical oversight and concurrence, and Readiness Review quality assurance personnel provided QA surveillance of l the review activities. Statements from the technical consultant and QA regarding their involvement and conclusions reached are ,

provided in section 8 of this module.

A brief summary of the two reviews and the method used in classifying findings resulting from the reviews are provided below.

Findino Classification  !

Following evaluation, findings were subjected to categorization as follows to indicate their relative importance:

Level 1- Violation of licensing commitments, project procedures, or engineering requirements with indication of safety concerns.

f' Level II - Violation of licensing commitments or

engineering requirements with no safety concerns.

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4 Level III - Violation of project procedures with no safety concernc.

Design Program Verification The NSSS for the Vogtle Electric Generating Plant (VEGP) is supplied by Westinghouse. The design of the NSSS for VEGP is not unique, but is generic for several Westinghouse-supplied NSSSs, including several in commercial operation. Westinghouse designed the NSSS using functional groups who use the same procedures and policies to perform the same technical functions for each Westinghouse NSSS. These functional groups work under the Westinghouse QA program, which has been accepted by the USNRC for generic application. In addition, Vogtle-specific QA audits have verified that the QA program for Plant Vogtle has been properly complied with.

Because of the evidence of adequacy of the Westinghouse design program, established through numerous reviews by the NRC as well as numerous audits by Westinghouse customers, including GPC, the design verification scope concentrated on a review of the NSSS design interface (i.e., the flow of design information between Westinghouse and Bechtel) which provides the link between the Westinghouse-generic design and VEGP-specific design requirements.

Eleven key areas of the NSSS design interface between Bechtel and Westinghouse, such as pipe stress analysis and accident analysis, were reviewed to ascertain whether the interface activities were performed in a controlled and effective manner.

The aspects considered were information transmittals requitad by each organization and receipt, correct internal distribution, and implementation of the information. The design program verification also ascertained whether the NSSS-related licensing commitments considered unique to VEGP had been incorporated into Bechtel project design criteria and other implementing design documents.

This verification resulted in four findings. One of these was classified as Level III, two as Level II, and one as Level I.

The four findings were: l

1. Finding 16-11 (Level II) resulted from a review of 10 CFR 50.55(e) reports, issued by GPC, associated with several Westinghouse-supplied components. In one report, commitments had been made to replace potentially defective pinion keys in certain Limitorque motor operators. The lh commitment to replace the pinion keys had not been followed through, nor was an effective tracking system in place to ensure proper followthrough of this commitment. The Project reviewed additional 10 CPR 50.55(e) reports and found that appropriate actions had been taken. The case of the pinion keys was considered to be an isolated oversight. The lh v4

Project has taken acceptable actions to upgrade their tracking system to include all past commitments made via f'N 10 CFR 50.55(e) reports to ensure that commitments in N- 10 CFR 50.55(e) reports are properly tracked and implemented.

2. Finding 16-12 (Level III) involved inadequate documentation, in BPC pipe stress calculations, of the acceptability of calculated dynamic acceleration levels applied to

(} Westinghouse-supplied valves. Within several pipe stress calculations only a reference was made to the project Design Criteria No. DC-1017, Stress Analyses Criteria, Rev. 3, for the allowable accelerations. No numerical comparison was

> documented in the body of the calculation. In addition one

' () case was identified in which the Design Criteria No. DC-1017 did not contain the allowable acceleration levels for one type of Westinghouse supplied valve, even though the calculation had made reference to the Design Criteria Document. Additional project reviews determined that for all other valves requiring seismic qualification the vendor-supplied.allowables are appropriately tabulated in DC-1017. This criteria has been revised to incorporate the missing allowables for the one valve identified above.

Regarding the lack of appropriate documentation of comparing actual versus allowable accelerations in pipe stress calculations, it was determined that the problem exists for calculations performed prior to June 1982, at which time the

/' calculation format was revised to include the appropriate k comparison. Proper documentation of the calculated versus allowable dynamic acceleration comparison will be i incorporated into the stress calculations (pre-1982) during the piping as-built reconciliation activities prior to fuel load.

3. Finding 16-13 (Level I) identified instances in which the instrument installation drawings issued by Bechtel did not fully incorporate Westinghouse installation requirements.

Project review determined that this problem was due to the method in which installation details were issued for instrumentation and to lack of review of vendor requirements t

fs prior to issuing those details. It was determined that a

( total of 17 instruments had been installed using incorrect bolt size and 47 had been installed using incorrect torquing requirements as a result of this discrepancy. This finding was evaluated for potential reportability and was determined not to be reportable. However, the instruments in question were remounted in accordance with vendor requirements.

((-)g Acceptable corrective action has also been taken in correcting the installation drawings and cross-referencing applicable vendor documents on the installation drawings to ensure that vendor installation requirements will be complied with for future instrument installations.

4. Finding 16-15 (Level II) stated that project procedures do not adequately outline the verification process for seismic qualification of Westinghouse NSSS equipment. The role vii

of Westinghouse, Bechtel, and others involved in the verification process is not defined by procedures. It could not be established by Readiness Review that a complete ll program exists for verification of NSSS equipment seismic qualifications (e.g., as qualified versus actual seismic levels). BPC has committed to expand the equipment qualification procedure to more completely address the seismic quailification verification process, for NSSS equipment, by December 1985. llh In ccnjunction with effective implementation of corrective actions committed to by the Project in response to the identified findings, the design program verification results indicate that the Bechtel/ Westinghouse interfaces are adequate.

The identified findings do not constitute a trend which could adversely affect the adequacy of the NSSS interface controls lh between Bechtel and Westinghouse or the associated design activities.

Details of the design program verification are presented in section 6.1.

Construction Program Verification The construction program verification consisted of commitment implementation assessment and construction assessment.

Commitment implementation assessment determined whether construction incorporated licensing commitments into the lll implementing documents, whereas construction assessment determined whether the NSSS component installation met the design requirements.

During commitment implementation assessment, 11 licensing commitments were identified as the responsibility of construction for implementation. The Readiness Review construction team identified approved project documents utilized by construction that invoked, by reference or detailed directions, the requirements of each commitment.

During construction assessment, approximately 50 hardware 3 elements and 800 records were assessed for compliance with the W appropriate drawing and specification requirements.

This assessment resulted in five findings (16-6, 16-7, 16-8, 16-9, and 16-10). Of the five findings, none were categorized as Level I, three (16-6, 16-8, and 16-10) were categorized as Level II, and two (16-7 and 16-9) were categorized as Level III. Of those five findings, 16-8 involved incomplete code stamping of a component; 16-6, 16-7, 16-9, and 16-10 were documentation errors. Corrective action taken by the Project on each was adequate to correct the errors and preclude recurrence. None of the findings brought into question the acceptability of the components.

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I The NISCO installation program was found to be well-organized I and controlled. The QA records were determined to be p(_)s identifiable, retrievable, and effective at demonstrating the acceptability of the hardware installations.

Details of the construction program verification are presented in section 6.2.

Readiness Review Team Conclusion i

Having performed a review of project documentation and the l primary loop components, Readiness Review concludes that  !

adequate controls exist to ensure the quality of work and the implementation of licensing commitments within the scope of this

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Based on the results of the review and implementation of effective corrective actions committed to by the Project, it is the conclusion of the Readiness Review Team that the design and construction programs and processes associated with the NSSS, within the scope identified in this module, will produce a final product that meets licensing commitments and design requirements.

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TABLE OF CONTENTS O Section Number Title 1.0 Introduction

() 1.1 1.2 Scope Module Organization 1.3 Vogtle Project Status 2.0 Organization and Division of Responsibility 2.1 Design Organization 2.1.1 Bechtel Organization 2.1.2 Westinghouse Organization 2.1.3 Bechtel-Westinghouse NSSS Interface 2.2 Construction 2.2.1 Georgia Power Company - Nuclear Construction 2.2.2 NISCO 2.3 Nuclear Installation Services Company Training and

's Qualification

{J i 3.0 Commitments 3.1 Introduction 3.2 Definitions l 3.3 Sources 3.4 Commitment Matrix 3.5 Implementation Matrix 4.0 Work Activities

() 4.1 Design 4.1.1 Interface Control 4.1.2 Implementation Matrix 4.2 Equipment and Materials

() 4.3 Material Control 4.3.1 Requisitioning 4.3.2 Receipt Inspection 4.3.3 Document Review Acceptance 4.3.4 Material Traceability I xi l

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I TABLE OF CONTENTS Section Number Title j 4.4 Fabrication, Installation, Inspection, and Testing I 4.4.1 Introduction 4.4.2 NSSS Components and Supports 4.5 Turnover to GPC i 1

5.0 Audits 4

5.1 Project Organization Audits 5.1.1 Audits of Design Activities 5.1.2 Audits of Construction Activities 5.2 Nuclear Regulatory Commission Inspections 5.2.1 NRC Inspections - Design 5.2.2 NRC Inspections - Construction 5.3 Past Construction and Design Problems 5.3.1 Design 5.3.2 Construction 5.4 Supplemental Audits 5.4.1 Authorized Nuclear Inspection Agency Audits 5.4.2 Westinghouse Audits 5.4.3 NISCO Internal Audits l

6.0 Program Verification 6.1 Design Program Verification 6.1.1 Summary 6.1.2 Verification Scope and Plan 6.1.3 Verification Review 6.1.4 Design Program Verification Findings i 6.1.5 Findings Significance 6.2 Construction Program Verification  !

6.2.1 Summary Evaluation 6.2.2 Prooran Assessment Plan 6.2.3 Commitment Implementation l xii

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Section Number Title 7.0 Independent Design Review i

() 8.0 Program Assessments / Conclusion 8.1 Summary of Open Corrective Actions l 8.1.1 Section 6.1 l 8.1.2 Section 6.2 l

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8.2 QA Statement 8.3 Technical Consultant's Statement 8.4 Readiness Review Board Statement 8.5 Engineering and Construction Management Statements 8.6 Resumes O

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1.0 INTRODUCTION

O l 1.1 SCOPE l l

The review documented in this module was conducted to P.scertain l whether design and construction work activities associated with t

'() specific aspects of the nuclear steam supply system (NSSS) comply with licensing commitments and whether compliance with these commitments is verifiable with existing project documentation.

Within the design area, this module addresses the design

() interface between Bechtel Power Corporation (BPC) and Westinghouse. Specific BPC design activities regarding the Westinghouse-supplied NSSS systems are addressed in other. '

modules (Table 1.1-1). Work activities considered Westinghouse generic are not addressed; however, this module addresses those Westinghouse activities considered Vogtle specific. Within the l construction area this module addresses those activities involved-with the installation of primary loop equipment only.

Installation of other Westinghouse-supplied NSSS hardware is I addressed in other modules (Table 1.1-1).

1 The effective date of this module is June 1, 1985.- That is, l

-changes in the included programs, organizations, commitments, j etc., occurring after this date are not addressed. )

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TABLE 1.1-1 MODULE HARDWARE / PROGRAMS Module 4 - Mechanical Equipment and Pipinq o ASME Section III Components O - Pumps (including drivers) r

- Valves (manual and power actuated)

- Heat exchangers

- Piping (including the installation program for the primary loop)

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Module 8 - Structural Steel o Pipe whip restraints (design only)

Module 11 - Pipe Supports o Pipe supports o Stress analysis o N-stamp program o Piping system as-built program o Pipe whip restraints (installation)

Module 16 - NSSS o Primary loop components

- Reactor pressure vessel

- Steam generator

- Reactor coolant pump

- Pressurizer

- Primary loop piping (design interfaces only)

- Etc.

Module 20 - Instrumentation and Controls o Instrumentation and controls O

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1.2 MODULE ORGANIZATION This module is divided into the following sections:

1. Introduction

2. Organization and Divisica of Responsibility - A brief

[ description of the project organizations and division The of responsibilities as they apply to this module.

overall project organization is discussed in Appendix A - Organization.

3. Commitments - This section contains project licensing commitments pertaining to the nuclear steam supply

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system within the scope of this module, as found in the FSAR, generic letters, and other documents. This section also lists documents that demonstrate implementation of these commitment..

4. Work Activities - A brief description of the processes for design, procurement, and construction applicable to the scope of this module.

5. Audits - A description of the level of audit activity by the NRC and various quality assurance organizations associated with the Vogtle Project as it applies to 7- this module. Also included in this section is a

'q"'3/ description of special investigations performed on work discussed in this module and previously identified problems.

6. Program Verification - A description of the verification plan development, implementation, and results, including corrective actions.

7. Independent Design Review - Provided as a separate report.

8. Assessment - Evaluations and conclusion of the subject work by the VEGP Readiness Review Board module expert,

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Readiness Review program quality assurance staff, s

Readiness Review Board, Engineering management and Construction management. In addition, this section contains a listing of finding items (section 6 of this module) still open and requiring project resolution.

~T Resumes of Readiness Review Team members involved in the development of this module are also provided.

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1 1.3 VOGTLE PROJECT STATUS i'3

1. The Vogtle Project was reactivated in the spring of 1976 after a

2 year shutdown. Following reactivation, the architect-engineer, Bechtel Power Corporation, rebid mechanical equipment and piping l contracts with the various suppliers, with the exceptica of the I nuclear steam supply system which in supplied by Westinghouse.

O The nuclear steam supply system equipment installer, Nuclear l Installation Services Company, arrived at the Vogtle )

l construction site in August 1982. I As of June 7, 1985 the approximate project status for Unit 1 and I

() common systems within the scope of this module is as follows:

Percent Complete i

Design 94 Construction 85 )

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l 2.0 ORGANIZATION AND DIVISION OF RESPONSIBILITY C

Georgia Power Company (GPC), acting on its own behalf and as agent for the Oglethorpe Power Corporation, the Municipal Electric Authority of Georgia, and the City of Dalton, is responsible for the design, procurement, and construction of the Vogtle Electric Generating Plant (VEGP).

The Western Power Division of Bechtel Power Corporation (BPC) is contracted by GPC to provide architect / engineering services.

BPC is the N-certificate holder for piping systems installed by s Pullman Power Products (PPP), except portions of the nuclear steam supply system (NSSS) for which Westinghouse Electric Corporation (Westinghouse) has the N-certificate responsibility. When BPC acts as designee for GPC, BPC may perform activities assigned to GPC where specifically permitted by ASME section III, division 1.

Georgia Power Company is the owner and, as an approved material

. lier authorized by BPC and Westinghouse, purchases and ,

eo rols ASME items and provides onsite construction services I to- calibration, maintenance, documentation, document control, l and nonconformance control.

Pullman Power Products is the installer (NA certificate holder)

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for ASME section III, division 1 piping systems and controls the scope of work for installation, field fabrication, operations, I examinations, tests, and inspections. Those work activities l performed by PPP are further defined and discussed in Modules 4 and 11, and are excluded from this module.

Westinghouse Electric Corporation is the NSSS supplier.

Westinghouse Electric Corporation has N-stamp responsibility for 1 the reactor coolant system, bottom-mounted instrument tubing, reactor vessel head vent system and 0-ring seal leak detection leak-off appurtenance, pressurizer safety and relief valve assembly, and pressurizer surge line, with pending responsibilities for the resistance temperature detector bypass manifold.

Hartford Steam Boiler Inspection and Insurance Company is the authorized nuclear inspection agency (ANIA) contracted to perform the ASME Boiler and Pressure Vessel Code-required inspections at the VEGP jobsite.

() Nuclear Installation Services Company is an installer (NA certificate holder) for ASME section III, division 1, class 1, 2, 3, and CS components, appurtenances, piping subassemblies, supports, and MC components and NF supports under the Westinghouse scope of supply.

() This section includes a brief description of the organization and responsibilities of Georgia Power Company, Bechtel Power 6

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Corporation Westinghouse Electric Corporation, and Nuclear Installation Services Company as related to the scope of this module.

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2.1 DESIGN ORGANIZATION The two design organizations with responsibilities within the scope of this module are Bechtel Power Corporation and Westinghouse Electric Corporation.

G 2.1.1 BECHTEL ORGANIZATION The Bechtel Power Corporation employs the matrix organization concept with an individual assigned as project engineering manager (PEM) and functional group heads reporting to him for the performance of functional tasks. Project direction is 9 provided by the PEM, while functional direction is provided by discipline chief engineers.

The Bechtel engineering design effort involving data which requires interface with Westinghouse, the nuclear steam supply system (NSSS) vendor, is performed by the various technical disciplines. The nuclear discipline has responsibility for NSSS interface administration, but civil-structural, control systems, electrical, mechanical, and plant design disciplines from both Home Office Engineering (HOE) and Project Field Engineering (PFE) are also involved with the NSSS interface. Each discipline is supervised by an engineering group supervisor (EGS) who provides technical direction and guidance for the lh respective discipline. The discipline work is divided into several activities which are supervised by engineering group leaders reporting to the EGS. The details of the EGS's role and responsibilities are provided in Appendix A.

The organization of the various Bechtel disciplines is described in detail in the modules covering their primary areas of responsibilities, such as Modules 1, 4, 17, and 20.

2.1.2 WESTINGHOUSE ORGANIZATION The Vogtle NSSS vendor is the Westinghouse Electric Corporation Water Reactor Divisions. The Water Reactor Divisions consist of ew) q_ several divisions, including the Electromechanical Division (EMD). Energy Systems Service Division (ESSD), Nuclear Components Division (NCD), Nuclear Fuel Division (NFD), Nuclear Operation Division (NOD), Nuclear Technology Division (NTD), and Plant Engineering Division (PED). The Vogtle NSSS is managed (s within Westinghouse by the project manager, Southern Company

(,) projects, NOD. The NSSS functions within Westinghouse report to this project manager on a matrix basis. For site-related matters, the project manager, Vogtle site reports directly to this project manager. Fuel-related matters are the responsibility of the Nuclear Fuel Division project manager for rs Vogtle. This organization is illustrated in Figure 2.1-1.

l l Westinghouse is organized on a functional basis. Design activities are performed by functional groups which perform the same function for each Westinghouse NSSS. The procedures and lh practices employed are the same for each NSSS and are not unique to the Vogtle NSSS. The scope of the Vogtle NSSS is defined by the NSSS contract between Secrgia Power Company and Westinghouse l Electric Corporation. The Vogtle NSSS, like other Westinghouse NSSSs. is designed under the Westinghouse Quality Assurance l

' Program which is carried out in conformance to the quality lh assurance plan described in topical report WCAP 8370/7800. This plan has been submitted to and accepted by tne USNRC for generic application to all safety-related work conducted in conjunction with design of commercial nuclear power plant projects. The American Society of Mechanical Engineers (ASME) regularly reviews the Westinghouse program, evaluating the areas of design lh control, interface control, and related disciplines. As a result. Westinghouse holds an ASME N certificate of authorization.

2.1.3 BECHTEL WESTINGHOUSE NSSS INTERFACE l The primary interface for the NSSS is between the Bechtel project engineering manager, Plant Vogtle and the Westinghouse I project manager, Southern Company projects. An interface exists at the plant site for disposition of deviation reports between the Bechtel project engineer-field and the Westinghouse Vogtle site manager. These interfaces are controlled by procedures within each respective organization.

The Bechtel project engineer has overall responsibility for the Bechtel interface for the NSSS. The day-to-day responsibilities are delegated to an assistant project engineer (APE),

responsible for NSSS coordination. NSSS contract administration is assigned to the nuclear discipline. The NSSS engineering group leader (EGL) reports to the nuclear group supervisor on matters related to the NSSS nuclear discipline. The NSSS EGL also reports to the APE on the status of the NSSS interface. In the latter capacity, the NSSS EGL maintains logs necessary to l identify and coordinate the NSSS interface and:

o Develops and maintains guidelines for the handling and distribution of design information from Westinghouse:

o Ensures that design information is formally documented; o Monitors open technical items between Bechtel and Westinghouse and ensures that such issues are resolved and their resolution is documented.

This organization is shown in Figure 2.1-2.

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j PROJECT MANAGER NUCLEAR FUEL SOUTHERN COMPANY DIVISION PROJECTS I

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PROJECT MAN AGER VOGTLE StTE

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ELECTOMECHANICAL l ENERGY SYSTEMS  ! NUCLEAR COMPONENTS

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i DIVISION SERVICE DIVISION DIVISION l l l l

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L_ NUCLEAR OPERATIONS .

Ni)CLE AR TECHNOLOGY __

PLANT ENGINEERING 4

DIVISION DIVISION DIVISION O -

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Figure 2.1-1 Westinghouse Organization

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] ENGINEERING MANAGER

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ASSISTANT j PROJECT ENGINEER q 4

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' I ENGINEERING GROUP g SUPERVISOR NUCLEAR DISCIPLINE , l

Q , ENGINEERING GROUP l I I l LEADER NSSS --J NUCLEAR DISCIPLINE

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CIVIL /STRU'CT UR AL l CONTROLS l ELECTRICAL MECHANICAL DISCIPLINE DISCIPLINE DISCIPLINE DISCIPLINE l l

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! NUCLEAR PLANT DESIGN DISCl?LINE DISCIPLINE i

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O Figure 2.1-2 Bechtel NSSS' Organization 4

i 2.2 CONSTRUCTION l l

Georgia Power Company (GPC) Nuclear Construction has j responsibility for construction of Plant Vogtle. For direct l work, Nuclear Construction employs Nuclear Installation Services j Company (NISCO) for the installation of nuclear steam supply L system (NSSS) equipment supports and installation and assembly of the NSSS equipment.

This section describes the organizations responsible for activities in the scope of this module.

O,j 2.2.1 GEORGIA POWER COMPANY - NUCLEAR CONSTRUCTION GPC Nuclear Construction provides direction to the contractors on thn schedule and budget, supplies all materials, provides

! drawings and specifications developed by Project Engineering.

l and performs the surveillance of construction activities.

l i These activities are provided through sections of the Nuclear l Construction Department and are described below. Appendix A d7 scribes the current Nuclear Construction Department l

organization.

2.2.1.1 Mechanical Project Section

}

The Georgia Power Company Mechanical Project Section provides coordination and support for contractors. This includes providing assistance in the following areas:

o Developing mechanical construction procedures and assuring they are in compliance with Project Engineering specifications and any applicable codes; o Resolving problems regarding techanical work including constructability issues, deviation reports, trends, field change requests, and open items; o Dispositioning deviation reports and open items; o Providing material for the contractors by initiating purchase orders and releases as required:

o Providing schedule and budget input to various site

,-) organizations; i

o Interfacing extensively with Coordination and Quality Control (QC) Groups on problem identification and resolution.

O

2.2.1.2 GPC Field Construction Operations - Coordination The Site Coordination Group directs work at Plant Vogtle and ensures work is completed on time. They interface with the site contractors to facilitar, work flow. The lower tier coordination groups help bring field conflicts and problems to the attention of the area engineers and inform QC when inspection hold points are reached. They maintain a watch for g productivity and quality problems. The Site Coordination Group is responsible for survey and layout work on the project.

The Site Coordination Group reports to the building manager.

2.2.1.3 Warehouse The warehouse is responsible for the receipt, storage, and issue of materials, components, and equipment for the project.

The warehouse personnel, along with GPC QC, ensure that correct materials are received and issued to the contractors.

2.2.1.4 GPC Mechanical Quality Control Section GPC Mechanical QC is responsible for performing receiving inspections, storage inspections, and inspections of equipment maintenance.

Additionally, Mechanical QC performs a random review of the radiographic film produced by the contractors.

Quality Control personnel performing inspections or reviews are certified to ANSI N45.2.6 and / or SNT- TC- 1 A.

2.2.1.5 GPC - Quality Lentrol Surveillance Section The QC Surveillance Section performs an overview of contractor QC activities. These overviews are performed at random and consist of:

o Daily observations of the contractor craft and QC inspectors for compliance with specifications and procedures, '

o Sampling of completed work that has received final inspection by contractor QC to ascertain compliance g; l with drawings, specifications, and procedures. '

The GPC surveillance inspectors are certified to ANSI N45.2.6 and/or SNT-TC-1A requirements and report areas of noncompliance to Project Engineering on a Deviation Report.

2.2-2

i l

2.2.1.6 GPC Drawino Control Section GPC Drawing Control provides drawings and specifications to the contractors and GPC sections.

l Drawing Control receives copies of drawings and specifications l l g- from Project Engineering and makes distribution to affected

', holders by way of a controlled distribution list.

l Drawing Control is also responsible for issuing change documents l and notifying recipients of any documents that have been voided or superseded.

l() Additionally, Drawing Control audits the contractors and GPC sections drawing control to ensure proper control and handling is maintained.

2.2.1.7 GPC Document Review Section  !

I GPC Document Review reviews documentation provided upon receipt from the vendors and determines compliance with the purchase o-der and specification. Upon acceptance, Document Review l allows issue of vendor-supplied materials, components, or l equipment.

/~ Additionally, Document Review is responsible for maintaining

( quality assurance documentation, documentation supplied by other l GPC sections and vendors, and documents turned over to GPC by )

l the contractors.

Document Review is discussed in more detail in Appendix E.

2.2.2 NISCO The NISCO site organization is presented in Figure 2.2-1. The responsibilities and duties of key position personnel are I described as follows:

2.2.2.1 Site Manager 4

I The NISCO site manager reports to the NISCO project manager. He is directly responsible for all production activities at the jobsite. His duties include:

O o Reviewing and approving field purchase requisitions; o Providing direction to site personnel; o

l()

i Arranging transfer of quality records to GPC upon completion of the project; 2.2-3

_ ~, - - _ -

1 o Assuring compliance of production personnel with the NISCO QA manual, including implementing any necessary changes resulting from QA, ANIA, or ASME audits; o Liaison with GPC site representatives.

2.2.2.2 Lead Engineer g

The lead engineer reports directly to the site manager and is responsible for the performance of ASME Code work on the project. He receives technical direction from the manager, project engineering.

2.2.2.3 Superintendents O

Superintendents report directly to the site manager and are responsible f or production wor:t performed by the crafts, and for

, compliance with the Process Control Sheets.

2.2.2.4 Field OA/OC Manager The field QA/QC manager reports to the manager, quality assurance and is responsible for all field QA/QC activities. He is assisted in his duties by the field QC engineers. His responsibilities include:

o Supervising all QA/QC activities; o Directing the gathering, filing, and indexing of QA records at the site; o Reviewing and approving Process Control Sheets; o Reviewing Process Control Sheets and drawings with the authorized nuclear inspector (ANI) for establishment of hold points; o Reviewing and approving heat treatment charts and making them available for review by the ANI; o Assuring that required personnel qualification records are at the jobsite and procedures are qualified, as required by ASME Code; o Supervising and monitoring of nondestructive examination (NDE) services subcontractors and heat treatment subcontractors, as applicable; o Review and approval of material certifications and data reports, and submitting them to the ANI:

2.2-4

1 1

o Liaison with the ANI for all jobsite activities: I o Coordinating QC activities with the site manager to assure that all QC requirements are properly scheduled and met; j

o Presenting proposed Use-As-Is and Repair nonconformance p)g s, dispositions to the ANI for ASME code items; o Providing monthly QA program status to the manager, quality assurance:

\

o Ensuring proper control and correction of nonconforming s conditions; o Approval of NISCO and vendor quality documents relating to site activities; o Maintaining vendor performance file; j o Distribution of the approved vendors list (AVL).

i Y

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0003p/323-5 1

2.2-5 i

h h h A_ 1

m. . .

TO CORPOHATE vtCE PRESIDENT TO COPORATE MANAGER OF CONSTRUCTION QUAlfTV ASSURANCE S,,e u. NAGER _____________._______

Lo O~OC MANAGER p-__________.

I I- ,

8 I

E

' OC ENGINEERS OC ENGINEERS I

OC OPERATIONS NDE OPERATIONS I

I I

SUPE R*N TENDEN T SUPERINTENDENT  ! OFFICE MANAGE R ENG!NEER l 1

I I

I I

ASSISTANT DOCUMENT FIE LD FNGINEE R3 SECRETARY PAv00tLCLERK SP RMMM 7gg l

CRAff 8:LLING CLERM PAYROLL CLERK I LEGEND ASSIST AN T ASStSTANT REPORTING RESPONS:BitITY CLERM CLERK

--- - - -- - COMuuNICA TION so.

'tNTE RFACE WITH CORPORATE PURCHASING AGENT Figure 2.2-1 Nuclear Installation Services Company Site Organization

2.3 NUCLEAR INSTALLATION SERVICES COMPANY TRAINING AND

() OUALIFICATION l Nuclear Installation Services Company (NICSO) staff and i supervisory personnel at Plant Vogtle are indoctrinated to the l requirements of the NISCO quality assurence program requirements l g in accordance with NISCO procedure ES-118, Quality

) Indoctrination Program, which is administered by the field QA/QC manager. Specific personnel applicability and the minimum subject matter requirements for each are as follows:

o Site manager

() -QA manual: all sections.

-Engineering specifications: all.

-Welding procedure specifications: all; o Lead engineer

-QA manual: all sections,

-Engineering specifications: all,

-Welding procedure specifications: all; j o Field engineers

-QA manual: Sections 2.0, 5.0, 6.0, 7.0, 9.0, 10.0, 11.0, 14.0, and 17.0,

-Engineering specifications: as applicable.

l 1 -Welding procedure specifications: as applicable; o Field QA/QC manager l

-QA manual: all sections,

() -Engineering specifications: all,

-Welding procedure specifications: all; o QA/QC engineers l

l -QA Manual: sections 1.0, 3.0, 4.0, 5.0, 6.0, 7.0, 8.0, 9.0, 12.0, 13.0, 14.0, 15.0, 16.0 and 17.0,

-Engineering specifications: all,

-Welding procedure specifications: all; 1

o Documentation clerks

-QA manual: sections 3.0, 12.0, and 17.0,

-Engineering specifications: as applicable.

-Welding procedure specifications: none; o Staff purchasing agents

-QA manual: sections 4.0 and 17.0,

-Engineering specifications: as applicable,

-Welding procedure specifications: none; o Craft supervisor I

-QA manual: sections 1.0, 3.0, 5.0, 6.0, 7.0, 11.0, 14.0, and 17.0,

-Engineering specifications: as applicable,

-Welding procedure specifications: all; o Craft foreman

-QA manual - Section 5.0, 6.0, 7.0, 11.0, 14.0, and 17.0, h

-Engineering specifications: as applicable,

-Welding procedures specifications: as applicable; o Craft personnel

-QA introduction letter,

-Engineering specifications: as applicable.

-Welding procedure specifications: as applicable.

-Quality memos: as applicable.

-Job requirements.

Field QC engineers are also trained to the requirements of ES-116-1, Qualification and Certification of Nondestructive Examination Personnel, and ES-ll6-2, Qualification and l

Certification of Inspection Personnel, to ANSI N45.2.6, as applicable for their respective areas of activity.

O 2.3-2

Training and certification of NDE personnel to ES-ll6-1 complies

,j with the requirements of:

o ASNT-SNT-TC-1A (1975);

o AWS-Dl.1; o ASME Boiler and Pressure Vessel Code section III, Division 1; o ASME Boiler and Pressure Vessel Code section V; )

i o ASME Boiler and Pressure Vessel Code section VIII;  ;

o ANSI Power Piping Code B31.1.

Training and certification of other QA/QC inspection p3rsonnel to ES-ll6-2 is in compliance with ANSI N45.2.6 (1978). NISCO procedure ES-117 Inspection, Testing, and Examination Personnel Training, details the recommended subject matter, length of l study, and texts to be used in the training of personnel under l the requirem 7ts of ES-116-1 and ES-116-2.

QA auditors are' trained and certified to NISCO procedure ES-116-3, which complies with the requirements of ANSI standards N45.2, N45.2.10, N45.2.12, and N45.2.23.

O O

O 0004p/323-5 2.3-3

3.0 COMMITMENTS

3.1 INTRODUCTION

This section contains, in matrix form, licensing and project commitments and the corresponding implementing documents. These are presented in two matrixes, the commitment matrix and the implementation matrix. A brief explanation of the development process for each matrix is also included.

Any differences between the commitments discussed in this section and the Vogtle Electric Generating Plant Final Safety Analysis Report, if any, are accidental, and the FSAR prevails.

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0 0065p/323-5 l

3.2 DEFINITIONS O Commitments are defined as the project obligations to regulatory guides, industry standards, branch technical positions, and other licensing requirements to the extent defined in the Final Safety Analysis Report.

An implementing document is that working level document that identifies project commitments as they apply to the specific work activity.

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O O 0082p/323-5

.. -- _ _ _ . - . , _. .~ _,. ,. _ ._.-__-.. , ,. , _ _ , _ . _ - .

3.3 SOURCES

' Commitments covered by this report are identified from the following sources:

o Final Safety Analysis Report including responses to Nuclear Regulatory Commission questions; O Responses to Generic Letters; o Responses to I&E bulletins.

These sources are reviewed for commitments based upon guidelines

() developed from the definition.

Implementation of commitments stated in the commitment matrix is typically contained in:

o Design criteria; o Equipment / material specifications; o Construction specifications; o Construction proccdures; o Technical specifications:

U( N o Operations procedures.

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O 0083p/323 5 l

3.4 COMMITMENT MATRIX Once identified by the Readiness Review team, the commitments are placed in the commitment matrix. Information identifying the source, source section, subject, and module are also indicated in the matrix. Any relevant comments concerning the commitments or subject of the section 7.0 indicated in the Remarks column.

The commitments included in this sectioa address the Vogtle-specific commitments. Commitments pertaining to the nuclear steam supply system (NSSS) are considered generic because of the functional organization of the NSSS supplier, Westinghouse. These commitments are not included in the commitment matrix because of their generic nature and the frequent and detailed level of review that Westinghouse receives from the Nuclear Regualtory Commission.

I 0063p/323-5

O O Q O O. O O 001941 TENTS SORTED BY SOURCE A10 SECTION (Df941 TENT C00911 TENT CopellTENT DOCUENT/ KSPONSIBILITY SOURCE SECTION SUBJECT FEATUK MODULE DESIGN CONST EMAf4CS EF NO.

EXPLANATION OF FIELDS COP 911 TENT SOURCE - The document containing the corvnitment (FSAR, Generic Letter, l.E. Bulletin Response, etc.)

CDP 911 TENT SECTION - Identifies the FSAR section, letter nun 6er, or question nurnber (DP94tTENT SUBJECT - % sub,iect of N FSAR section or Generic Letter l

DOCUMENT / FEATURE - 1he document discussed in the FSAR section or the plant feature described in the FSAR section MODULE - N Readiness Review modules applicable to the comitment under discussion ESPONSIBILITY - An X is placed under the heading for the organization responsible for inplenentation of the comitment i EF. NO. - A reference nurser that corresponds to the appropriate line entry In the ipplementation matrix i,

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? .___.. _____. _ _ .__ _ _ __.__._.______ _ ___ _ _.___ _ _ ~ _ _ _ _ . _ . _ _ _ _ _ _

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COMMITMENT COMMITMENT COMMITMENT DOCUMENT / MODULE RESPONS IBILITT REMARKS REF NO SOURCE SECTION SUBJECT FEATURE DESIGN CONST

==== ======== =e========= = = = = = = = = = = = = = = = = = = = = ==================== ====== ======= ======= ==================== ==

FSAR I. 9. 4 ASSUMPTIONS USED FOR RG 1.4, REv. 2, 6/74 16 I SEE 15.6.5-5 FOR 2021 EVALUATING THE CONFORMANCE POTENTIAL RADIATION CONSEQUENCES OF A LOCA.

FSAR 1. 9. 7 CONTROL OF RG 1.7 , REV. 2, 18A.16 I REF. 6.2.5.1.1 FOR 2041 COMBUSTIBLE GAS 11/78 21W CONFORMANCE.

CONCENTRATIONS IN CONTAINMENT FOLLOWING A LOCA FSAR I. 9. 29 SEISMIC DESIGN RG 1.29 REY. 3 04,16, I REF. TABLE 3.2.2-1 1532 CLASSIFICATION 9/78 11 FSAR 1. 9. 31 CONTROL OF FERRITE RG 1.31, REV. 3, 08.04, I SEE 5.2.3.4.6 SEE 1855 l CONTENT IN STAINLESS 04/78 16,18, NUMBER 256 FOR STEEL WELD METAL 11,14 IMPLEMENTATION FSAR 1. 9. 36 MON-METALLIC THERMAL RG 1.36 REV. O, 10,14. I REF. 5.2.3.2.3 & 1536 INSULATION FOR 2/73 16,21W 6.1.1.1.3 AUSTENITIC S.S. ,

FSAR 1. 9 37 QA-REQUIREMENTS FOR RG 1.37, WEV. O, 03,16, X REF. 5.2.3.4.1, 1537 CLEANING OF FLUID 'MARCR 73) 04 17.1.2, 17.2, RG.

SYS. & ASSOC. 1.88, RG. 1.33 & RG.

COMPONENTS 1.37 (3/73)

FSAR 1. 9. 37 QA-REQUIREMENTS FOR ANSI M45.2.1-1973 03,16, X X REF. 5.2.3.4.1, 1538 ,

CLEANING OF FLUID 04 17.1.2, 3.8.1, RG. }

SYSTEMS & ASSOC. 1.88 RG. 1.33 RG.  !

COMPONENTS 1.37, (3/73)

FSAR 1. 9. 37 OUALITY ASSURANCE R.G. 1.37, 3/73 03,16, I I CONFORMS FOR DESIGN 3201 REQUIREMENTS FOR 04 AND CONSTRUCTION-CAP CLEANING OF FLUID WITR EICEPTIONS AND SYSTEMS AND CLARIFICATIONS.

ASSOCIATED CONFORMS FOR COMPONENTS OF OPERATION OAP WITR WATER-COOLED NUCLEAR CLARIFICATIONS.

POWER PLANTS.

(1.9.37.2) t 1

6

_ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _._.____________.___._______________________________r - - - _ _ - . , = . -. -.m -_, m . . _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ _ - . _ _ _ _ .

~ .

Pcge Do. 2 11/19/85 COMMITMENTS i

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==== ==================== ==================== ================ === ====== ======= ======= ==================== ==

FSAR 1. 9. 37 QA-REQUIREMENTS FOR 10CFR50 APP. B 03,16 X REF. 5.2.3.4.1, & 4681 CLEANING OF FLUID 04 17.1.2 SECT. 17.2 SYSTEM AND RG. 1.88, RG 1.33 ASSOCIATED RG. 1.37 (3/73)

COMPONENTS FSAR 1. 9. 44 CONTROL OF USE OF RG 1.44 REY. 0.5/73 16.21W X REF. 5.2.3.4 1541 SENSITIZED 5. S.

FSAR I. 9. 45 RCPB LEAE DETE7 TION RG 1.45 REY. 0.5/73 16 I REF. 5.2.5 1542 SYSTEM FSAR 1. 9. 46 PROTECTION A1AINST BTP MEB 3-1 BTP ASB 04,16, X REF. TABLE 3.6.1-2 & 1543 PaPE BREAE INSIDE 3-1 21W SECT. 3.6 CONTAIN*A INT WCAP-8172-A, ANSI 20.2 DEAFT IN LIEU OF RG 1.46 FSAR 1. 9. 48 DESIGN LIMITS & RG 1.48 REV. O. 11,16, X REF. 3.9.N.3 & 1544 LOAblNG COMBIN. FOR 5/73 18,21J TABLES 3.9.N.3-3 SEISMIC CAT. I FLUID 21W TRROUGH 3.9.B.3-10 SYSTEM COMPONENTS FSAR l. 9. 50 CONTROL OF PREREAT RG 1.50 REV. O, 16.21W X REF. 5.2.3, 1546 TEMPS. FOR WELDING 5/73 WCAP-8577 0F LOW-ALLOY STEEL FSAR 1. 9. 53 APPLICATION OF RG 1.53 REY. O, 12.04, X REF. 7.1.2.6 & 1551 SINGLE-FAILURE 6/73 16,20, 15.0.8 CRITERION 06.18, 21W FSAR 1. 9. 61 DAMPING VALUES FOR RG 1.61 10/73 16.19, F 2188 SEISMIC DESIGN 21J,21 W

FSAR 1. 9. 82 SUMPS FOR EMERGENCY RG 1.82, 6/74 04,16 I SEE FSAR 6.2.2.2 711 CORE COOLING AND 07A CONTAINMENT SPRAY SYSTEMS O O O O O O O

O O O _ . _

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COMMITMENT COMMITMENT COMMITMENT DOCUMENT / MODULE RESPONS IRILITT REMARES REF NO SOURCE SECTION SUBJECT FEATURE DESIGN CONST

==== === =============== ==================== ==================== ====== ======= ======= ==================== ==

FSAR 1. 9. 84 DESIGN AND FAB. CODE RG 1.84, REV. 20, 18,08 I REF. TABLE 1.9-1, 1565 CASE ACCEPT. ASME 11/82 04,16, PARA. 5.2.1.2, TABLE III, DIV. I 11,20 1.9-3 FSAR I. 9. 85 MATERIAL CODE CASE RG 1.85, REV. 20, 08.18 I REF. TABLE 1.9-2, 1569 ACCEPT. ASME III, 11/82 04,16, 'l.9-3, PARA. 5.2.1.2 t

DIV. I

• 11.20 FSAR I. 9.116 0.4. REQUIREMENTS RG 1.116 REV. O, 03,04 I REF. CHAPTER 17, 1578 FOR INSTALLATION, 6/76 16,20 VEGP-QAP, RG's 1.58 INSPECTION & TESTING 05.14, 1.88, 1.74, 1.33, OF MECRANICAL 11 1.37, 1.38, 1.39.

EQUIPMENT AND SYSTEMS.

FSAR 1. 9.116. I Q.A. REQUIREMENTS ANSI M45.2.8-1975 03,04 I REF. CHAPTER 17, 1579 l FOR INSTALLATION, 16,20 VEGP-QAP, RG'S 1.58, INSPECTION & TESTING 1.88, 1.74, 1.33, OF MECMANICAL 1.37, 1.38, 1.39.

EQUIPMENT & SYSTEMS FSAR 1. 9.139 GUIDANCE FOR RRR RG 1.139, REY. O, 07A,03 I SECTION 5.4.7 1611 5/78 04,16 21W FSAR 3. 1. 4 CONFORM ANCE WITH NRC ICCFR50, APP. A. GDC 16,18A X 808 GDC, FLUID SYSTEMS 41 04,21 W

FSAR 3. 1. 4 NSSS COMPONENTS ANSI N18.2-1973 16.21W X SEE GDC 30 1727 i DESIGN CLASSIFICATION FSAR 3. 1. 4 CONTAINMENT RG 1.4 16 I REMOVAL OF 1740  !

j ATMOSPRERE CLEAMUI PARTICULATE FISSION PRODUCTS BY THE CSS FSAR 3. 2. 2 VEGP CLASSIFICATION RG 1.29 04.18, X 1754 SYSTEM 16,11, 21W 4

s 4

_ --w-. ----._,_.---,_----.-x_-..--_ . - - - _ _ _ - - - - - _ - - - - - - - - - - - - .--- ------,--,_-s_._-- _ - _ . - - - - - - - - - - - - - - - - - -__e.- --_-_' _a_am' ts.---w_ * --

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FSAR 3. 2. 2-1 CLASSIFICATION OF NOTE (t): REAT 16.21W X TABLE 2.2.2-1 847 '

STRUCTURES, EXCRANGER ASME III COMPONENTS AND STSTEMS FSAR 3. 2. 2-1 PRINCIPAL CODES AND TEMA-C OR TEMA-R 04,16, X TABLE 3.2.2-1, NOTE 859 STANDARDS FOR TUBULAR EXCR. MFR. 21W (g) 7.3.2.2-1 ASSOC. CLASS C OR R FSAR 3. 2. 2-1 PRINCIPAL CODFS AND ASME III, CLASS 04,08, X TABLE 3.2.2-1 880 STANDARDS FOR 1.2.3 OR MC, MF, OR 16.11 T.3.2.2-1 CS IR,21W FSAR 3, 2, 2-1 PRINCIPAL CODES AND NEMA MGl 1972 16,21W X 867 STANDARDS FOR TABLE (NATIONAL ELECTRICAL 3.2.2-1 MFRS. ASSOCIATION)

MOTORS / GENERATORS FSAR 3. 2. 2-2 CONST. CODES /STDS. ASME III, SUBSECTION 04.18, X TABLE 3.2.2-1 NOTE 885 Q.G.-B FOR PRESSURE NC, CLASS 2 16.08, (g) SUBSECT. NF FOR VESSELS, PIPING, 21W SUPPORTS, SUBSECT.

PUMPS, VALVES, ATM. NE, CLASS MC FOR STORAGE TANES, METAL CONT. COMP.,

0-15pois STORAGE SUBSECT. NC FOR CORE TANES, SUPPORTS, SUPPORT STRS. MSR METAL CONT. COMP, CORE SUPPORT STRS.

FSAR 3. 2. 2-2 CONST. CODES /STDS. ASME III. Sr1SECTION 16.21W X TABLE 1.2.2-2, QGA 888 Q.G. - A. FOR NB, CLASS 1 SUBSECfl0N NF FOR PRESSURE VESSELS, SUPPORTS PIPING. PUMPS, VALVES, SUPPORTS FSAR 3. 2. 2-2 CONST. CODES /STDS. ASME III, SUBSECTION 188,20 I TABLE 3.2.2-2, QGC 889

, 0.G. - C. FOR ND, CLASS 3 04,16 SUBSECTION NF FOR PRESSURE VESSELS, 18A,2 SUPPORTS PIPING. PUMPS, IN VALVES, ATM. STORAGE

( TANES, 0-15psig

[

STORAGE TAMES, SUPPORTS.

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1 PROPAGATION OF BREAK 04,16 X 18#,1 FSAR 3, 6. 2. 3. 2 BREAK PROPAGATION, LARGE RCS PIPING TO UNAFFECTED LOOPS PREVENTED j FSAR 3. 6. 2, 3. 2 BREAK PROPAGATION, PROPAGATION OF BREAK 04.16 X 1842 LARGE RCS PIPING IN AFFECTED LOOP DOES NOT EXCEED 20s OF FLON AREA 0F i

RUPTURED LINE FSAR 3. 6. 2. 3. 2 SMALL RCS BRANCR TOTAL BREAR AREA 16.21N X: 963 LINE BREAR LIMITED TO 12.5 S0.

PROPAGATION IN.

FSAR 3. 6. 2. 5. 2 PRIMARY & SECONDARY ASME III NB-3650 16 I 967 STRESS INTENSITT (NCAP-8082-P-A AND RANGES (RCL) WCAP-8172-A)

I FSAR 3. 8. 3. 2 CONTAINMENT INTERNAL ASME III. SUBSECTION 11.08, X X SEE OTRER 1997 STRUCTURES, CODES NF, 1977 INCLUDING 16.21N SUBSECTIONS OF FSAR AND STANDARDS WINTER 1977 ADDENDA 3.8.3 FOR MORE (COMPONENT SUPPORTS) SPECtJ1C DETAILS FSAP 3. R. 3. 4. 6 CONT. INT. ASMF TII, SUBSECTION 16 X. 4695 STRUCTURES BOLTS NF JTTACHING PRESSURIZER BASE TO STEEL SUPPORT FRAME FSAR 3. 8. 3. 6-1.G MATERIAL FOR STEEL ASME III, SUBSECTION 16 X 1785 LINEAR SUPPORTS OF NF 1 RCS FSAR 3. 9. B. 3. 1. 2 DESIGN AND NORMAL ASME III, NB-3112 & 21L,ll X 1658 CONDITIONS MB-3113 16 FSAR 3. 9.B. 3-1 LOAD COMBINATIONS LOAD COMBINATIONS 21J,11 X 1660 FOR ASME III, DIV. ,16 I, CODE CLASS 1, 2,

& 3 COMPONENTS AND SUPPORTS

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FSAR 3. 9.N. 1. 4.1 CRITICAL DAMPING FOR 24 AND 4% 16 x 4732 OBE & SSE SEISMIC RESPECTIVELY ANALYSIS FSAR 3. 9.N. 1. 4.1.C LOADING RCS SEISMIC ANALYSIS 16 X 4755 CONDITIONS-SEISMIC 6 STATISTICALLY INDEPENDENT TIME -

RISTORY INPUTS: 3 TRANSLATIONAL ACCELERATIONS. 3 ROTATIONAL ACCELERATIONS FSAR 3. 9.N. 1. 4.8.A STRESS CRITkRIA - LOADS COMBINED BY 16 X 4757 CL.1 COMP. L COMP. ALGEBRAIC SUM.

SUPPORTS - JESIGN FSAR 3. 9.N. 1. 4.8.B STRESS CRITERIA - DYNAMIC LOADS 16 X 4759 CL.1 COMP. & COMP. COMBINED USING SUPPORTS - NORMAL, SQUARE ROOT OF SUM UPSET OF SQUARES (SRSS)

FSAR 3. 9.N. 1. 4.8.C STRESS CRITERIA - NORMAL LOADS 16 X 4760 CL.1 COMP. & COMP. COMBINED SUPPORTS - EMERGENCY ALOEBRAI(ALLY.

DYNAMIC LOADS COMBINED BY SRSS FSAR 3. 9.N. 1. 4.8.D STRESS CRITERIA - LOCA & SSE LOAPS 16 I 4761 CL 1 COMP. & COMP. COMBINED BY SRSS ON SUPPORTS - FAULTED LOAD COMPONENT BASIS FSAR 3. 9.N. 1. 4.8.D STRESS CRITERIA - SUSTAINED LOADS (WT. 16 I 4762 CL.1 COMP. & COMP. EFFECTS) ARE SUPPORTS - FAULTED COMBINED WITH SRSS RESULTS BY ALGEBRAIC SUM FSAR 3. 9.N. 1. 4.8.D STRESS CRITERIA - OTHER DYNAMIC LOADS 16 X 4763 CL.1 COMP. & COMP. COMBINED USING SRSS.

SUPPORTS - FAULTED l

l O O O @ O O O

. l' Page No. 7 11f19/85 COMMITMENTS

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COMMITMENT COMMITMENT COMMITMENT DOCUMENT / MODULE RESPONS IBILITY REMARES REF NO SOURCE SECTION SUBJECT FEATURE DESIGN CONST

,========= ==e================= ==================== ====s=============== ====== ===e=== ======= ========e=========== ========

FSAR 3. 9.N. 3 ASME CLASS ASME III 11,16 I 4739 COMPONENTS '

7 FSAR 4.A. l.A CORE EXIT RG 1.75 16 I AMEND. 15 4951 TNE RM0 COUPLE -

MONITORING SYSTEM - t CABLE ROUTING FSAR 4.A. l.B CORE SUBC00 LING RG 1.75 16 II AMEND. 15 4952 i MARGIN MONITOR -  !

CABLE ROUTING FROM  !

SENSOR TO DISPLAT FSAR 5. 2. 1.1 COMPLIANCE WITN 10CFR50, SECTION 16.21W X 188 CODES AND CODE CASES 50.55a FSAR 5. 2. 3. 2.2 COMPATIBILITT WITN RG.I.36 16 X 211-EXTERNAL INSULATION AND ENVIRONMENTAL ATMOSPNERE FSAR 5. 2. 3. 3. 2 CONTROL OF WELDING ASME III & IX 16 I 4997 FOR FERRITIC MATERIAL FSAR 5. 2. 3. 4. 4 PREVENTION OF IGA .NO BLOCK WELDING .16 I 4998 ,

i MAX 350 DEGREE INTERPASS TEMP.

~

EXERCISE' WELD PROCED. APPROVAL FSAR 5. 2. 3. 4.6 NDE METRODS ASME III 21W,16 I 226

, FSAR 5. 2. 3. 4.6 WELD MATERIAL AND ASME III & IX 21W,16 X 223 i WELD PROCEDURE QUALIFICATION FOR AUSTENITIC S/S FSAR 5. 4. 14. 4 RC COMPONENT ASME V 21W,16 I MODIFIED ST ASME 315 SUPPORTS TESTS & III, NF INSPECTIONS ,

t s

___&_____._______.______ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ ______m. _ _ _ _ _ _ _ _ _ _ ___._._ __.____.__m___ __2_______.m.__ ___ ___._______2___ _ _ _ _ _ _ _ .

Page No. 8 11/19/85 COMMITMENT 5

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COMMITMENT COMMITMENT COMMITMENT DOCUMENT / MODULE RESPONS IBILITT REMARKS REF NO SOURCE SECTION SUBJECT FEATURE DESIGN CONST

====== ========= = ======== ==== == === == ======

l FSAR 6. 1. 1. 1 ESF MATERIAI-S ASME III. NC-2160 & 04.11 I 322 {

SELECTION & NC-3120 16.20 I I

FABRICATION 21W FSAR 6. 1. 1. 1 ESF MATERIALS SPEC. ASTM /ASME COMPLIANCE 04.11 X 326 REQUIREMENTS - NA/NCA-1220 16A.20 FSAR 6. 1. 1. 1 ESF CONST. MATERIAL 3 ASME III. NC-2000 & 16.04 X 327 CONTAINMENT NE-2000 20.21W FENETRATION MATERIALS FSAR 6. 1. 1. 1.1 MATERIAL SPECS FOR ASME III. NC-2000 04.11 I 324 RST COMPONENTS IN CLASS 2 16.20 QUALITT GROU9- B 18A.06 21W FSAR 6. 1. 1. 1.1 MATERIAL SPECS. FOR ASME III. ND-2000 04.11. I 325 ESF COMP. IN QUALITT CLASS 3 16.20 GROUP C 18A.21 W

i FSAR 6. 1. 1. 1.2 ESF CONSTRUCTION ASME II. PART C. SFA 04.11 I 328 MATERIALS- WELD 5.1 16.20.

MAT'L FOR JOINING 5.2.5.5.5.17.5.18 & 18A FERRITIC BASE MAT *LS 5.20 FSAR 6. 1. 1. 1.2 ESF CONST. ASME II. PART C, SFA 04.11. I 329 MAT *LS-MELD 5.11 & 5.14 16.20 MATERIALS FOR JOINING N1-CR-FE ALLOTS FSAR 6. 1. 1. 1.2 ESF CONST. MAT *LS ASME II, PART C. SFA 04.11 I 330 FOR JOINING 5.4 & 5.9 16.20 l AUSTENITIC S.S 21W

) 331 FSAR 6. 1. 1. 1.2 ESF CONST. MAT'LS ASME III & II 04.11 I WELD MATERIAL 16.20 QUALIFICATION 21J.18 A

G S S O e --

e e

O O O O O O O .

P1ge No. 9 11/19/R5 COMMITMENTS

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COMMITMFNT COMMITMENT COMMITMENT DOCUMENT / MODULE RESPONS IBILITY REMARKS REF NO I SOURCE SECTION SUBJECT FEATURE DESIGN CONST

===,== ========== ========= =e= ================ ==================== ====== ======= ======= ==================== ======

FSAN 6. 1. 1. 1.2 ESF CONST. MAT'L5 ASME II 04,16, X 332 AUSTENITIC S.S. 20,18A UTILIZED IN FINAL REAT TREATED COND.

F54R 6. 1. 1. 1.2 ESF CONST. LIMITED TO NO 16,20, X 333 MAT'L-COLD WOREED GREATER TRAN IBA  ;

AUST. 5.S. 90,000 psi YIELD STRENGTH FSAR 6. 1. 1. 1.2 ESF CONSTRUCTION COMPONENTS IN 04,16, X 2030 MATERIALS CONTACT WITR BORATED 21W WATER ARE FABRICATED OF OR CLAD WITR

-AUSTENITIC STAINLESS i STREL OR EQUIV.

CORROSION-RESISTANT-MATERIAL FSAR 6. 1. 1. I.3.A ESF MAT *LS INTEGRITY RG 1.31 04.08, X 256 0F SAFETY RELATED 20.16 COMPONENTS FSAR 6. 2. 2. 2 CSS SAFRTY DRSION CSS AUTOMATICALLY 16,20, X 2403 BASES PLACED-IN OPERATION 21W ON RECEIPT OF TWO OUT OF FOUN CONTAINMENT PRES 5URE

-(RION-3) SIGNAL 5 FSAR 6. 2. 2. 2 CSS SYSTEM IEEE 279 16,20, X 2406

, INSTRdMENT 21W APPLICATIONS FSAR 6. 2. 2. 2 DESIGN BASIS FOR SINGLE ACTIVE 16.04, X 2479 ,

CONTAINMENT SPRAT FAILURE PLUS LOSP 21W l (INJECTION) AND SINGLE ACTIVE OR PASSIVE FAILURE PLUS LOSP (RECIRCULATION) 4 4

. _ _ _ . - - _ _ - . - _ _ - . _ - _ . . - _ ~ _ - _ _ _ _ _ _ . - - _ - - - . - _ . _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ - - _ - _ _ _ - _ _ - - _ _ - - - _ _ _ _ - _ __._________--______-.-__---..-___-__-._________u . _ . _ _ - _ _ _ . _ . - - _ . . . . .

Page No. 10 11/19/85 COMMITMENTS

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COMMITMENT COMMITMENT COMMITMENT DOCUMENT / MODULE RESPONS IBILITY REMARKS REF NO SOURCE SECTION SUBJECT FEATURE DESIGN CONST

== = ==================== ==================== === ================ ====== ======= ======= ==================== ==

FSAR 6. 2. 2. 2 CONTAINMENT SPRAY ASME III, CLASS 2 16,04 X TALLE 3.2.2.l(P.6) 2480 SYSTEM CODE AND 21W STANDARDS FSAR 6. 2. 2. 2 CONTAINMENT SPRAY ASME Ill. CLASS 3 16,04, X TABLE 3.2.2.1(P.6) 2481 SYSTEM CODE AND 21W STANDARDS FSAR 6. 2. 2. 2 CONTAINMENT SPRAY NEMA-MG1 16,04, X TABLE 3.2.2.l(P.6) 2483 SYSTEM CODE AP 7 21W STANDARDS FSAR 6. 2. 2. 2 CONTAINMENT SUMP MINIMUM pH OF 8.5 16.04, X 2485 FLUID pH AT END OF 21W INJECTION PtASE FSAR 6. 2. 2. 2 MATERIAL TUiT CAN AUSTENITIC S.S. 16,04, X 2486 COME IN COfeTACT WITH 21W RECIRCULATION FLUID FSAR 6. 2. 2. 2 DESIGN OF TRE RG 1.82 16,04 X 2515 CONTAINMENT EMERGENCY CORE COOLING SYSTEM SUMPS FSAR 6. 2. 2. 2. 2. 3 NPSH AVAILABLE TO RG 1.1 16.04, X 2512 THE CONTAINMENT 21W SPRAY PUMPS FSAR 6. 2. 2. 2.1 CONTAINMENT SPRAY RG 1.32 16,21W X 2402 SYSTEM (CSS) SAFETY DESIGN BASES FSAR 6. 2. 2. 2.2 SPRAY ADDITIVE MAX SPRAY pH OF 10.5 16.21W X 2514 EDUCTORS FSAR 6. 2. 2. 2.3 CSS SAFETY DESIGNED TO CATEGORY 16.21W X 2404 EVALUATION I AND QUALITY GROUP B REQUIREMENTS O O O O O O O

( O O isga No. 'I II/19.H5 COMMITMENTS

,v===

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FSAR 7. I. 2. l.IO ESF MOTOR SPECS STARTING VOLTAGE AT 16 x 4484 LEAST 75% OF RATED VOLTAGE OF 4 KV (AND ABOVE) AND 460 VOLTS.

FSAR 9. 2. 8. 3 AUXILIARY COMPONENT RETURN LINE 16 x 4996 COOLING WATER SYSTEM AUTOMATICALLY

- RCP THERMAL ISOLATED IN CASE OF RARRIER A CRACK IN RCP THERMAL BARRIER FSAR 9. 3. 4. 1.3.5 LEAKAGE PROVISIONS CVCS COMPONENTS, 16.21W X 3455 YALVES & PIPING WHICH SEE RADI0 ACTIVE SERVICE ARE DESIGNED TO LIMIT LEAKAGE TO ATMOSPHERE.

FSAR 17. 1. 2 GPC QA PROGRAM RG. I.37 03.04, X 3361 REQUIRRMENTS: 16 I CLEANING OF FLUID SYSTRMS AND COMPONENTS

. I.E.B. C-80/08/05 VACUUM CONDITION LOW PRESSURE PROCESS 04,16, X RESPONSE TO IEB' 4674 j CORRES. RESULTING IN DAMAGE OR HOLDUP TAMES THAT 21W 80-05 TO CYCS HOLDUP CAN CONTAIN TANKS. RA10 ACTIVE MATERIAL ARE DESIGNED WITH l FEATURES TO PRECLUDE VACUUM CONDITIONS h I.E.B. C-81/ll/20 GATE-TYPE VALVE MAXIMUM PRESSURES 16 I RESPONSE TO IEB 4680 CORRES. CLOSURE AGAINST (psi) AS FLOW 81-02 DIFF. PRESS. APPROACRES ZERO (TABLE). {AFFECTED i YALVES WILL BE

! MODIFIED.)

1 1

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Page No. 12 11/19/85 1

COMMITMENTS l

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COMMITMENT COMMITMENT COMMITMENT DOCUMENT /

I SOURCE SECTION MODULE RESPONS IBILITY REMARKS REF NO SUBJECT FEATURE DESIGN CONST

=-- ===============,==== ========= ==3e ===== =re================= ====== ======= ======= ==================== ==

NRC QUEST. Q210, 46 STRESS LIMITS USED CORRES.

PRIMARY MEMBRANE IS 16 X Sm= DESIGN STRESS 4987 IN ANALYSIS OF VALVE < OR = 1.0 So INTENSITY PER ASME DISCS IN CLASS 1 PRIMARY MEMBRANE III. TABLES I-1.0 VALVES GREATER TRAN PLUS BENDING IS < OR 4 INCR = 1.5 Sm NRC QUEST. Q210. 46 STRESS LIMITS FOR PRIMARY MEMBRANE IS 16 CORRES.

X S = ALLOWABLE STRESS 4991 VALVE DISCS IN CLASS < OR = 1.0 S. VALUES PER ASME III, 2 AND 3 VALVES WHERE PRIMARY MEMBRANE TABLES 1-7.0 & I-8.0 ANALYSI' IS REQUIRED PLUS BENDING IS < OR

= 1.5 S NRC QUEST. Q4RO. 6 CONTA'NMENT DIFFERENTIAL 16 X 8'O R H E S . RESPONSE TO QUESTION 4153 FMERGd4CY SUMPS PRESSURE INDUCED BY DESic( CAPABILITY 50% BLOCKAGE WITH DEBRIS i

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=======,

MODUtg 16 - SORTED BY REFERENCE MUMBER

v=,= ========== ====e==========

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, s .- - : . = 1 -=== s= t==,,=.=2 ==2 == ===--_-=. ===.,= e == ,r =-==== == -======== ====== =-====, ======== ===e========= =-

t 10CFR$0 SECTION 16 DC-1201-2.0, DC-1201-2.0, 50.55m REV. 2, 3-9-84 HEV. 0,-5-11-78 RG 1.36 16 SEE REMARES SEE REF. NO.

1536 FOR DESIGN IMPLEMENTATIO N

ASMR III & IX 'l 6 NISCO PROCEDURES NISCO PROCEDURES E.S.-300, REV. E.S.-300, REY.

F, 4-8-83, C, 8-26-82 E.S.-GNP-BMI-1 E . S . -0W P- BM I- 1, REV. E, 1-14-85 REY. A. 6-7-83,

, 8- bb4 , b b b4 ASME 111 16 SEE REMARKS SEE REMARKS SEE REF. NO.

315 FOR CONSTRUCTION IMPLEMENTATIO N

RG l.31 16 DC-1206-1.0, . DC-1206-1.0, REV. 2, 4-28-83, REV. 2, 4-28-83, X4A001-10.3G, X4AQOI-10.3G, REY. 16, REY. 13 ..

7-20-84, 11-8-82, 14AQ10-6.4D.12, X4A010-6.4D.12, REV. 5. 12-17-84 REV. 2, 7-1-81 ASME V 16 NISCO NDE NISCO NDE PROCEDURES PROCEDURES

. E.S.-8.7 REV. E.S.-8.7,'REV. -

C, 8-23-84 A, 3-20-84 E.S.-100-1, REY. E.S.-100-1, REY.

A, 8-23-82, A, 8-23-82, E.S.-100-2, REV. E.S.-100-2, RrV.

E, 7-25-84, A, 8-25-82, E.S.-100-4-1, E.S.-100-4-1, REV. C, 6-21-84, REY. A. 6-21-84 E.S.-LOO-5, REY. E.S.-100-5, REY.

C, 3-9-83, C. 3-9 j E.S.-Il6-1, REV. E.S.-Il6-1, REV.

4 I, 11-3-82 H, 10-12-82 4

l

- --. . - - --- -- - - . - - . _ . . . . . - . . - - . - _ _ _ - _ = . . . -. ..

Page No. 2 11/19/85 IMPLEMENTATION

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====================-============

DOCUMENT / FEATURE SECTION MODULE DESIGN LAST DESIGN FIRST CONST LAST CONST FIRST REMARKS REF NO

========:-==== ============ ====== ================ ================ ================ ============== = =========== = ==

ASME III, NC-2160 & 16 DC-1204-2.0C.I. DC-1204-2.0, 322.00 NC-3120 REY. 2, 2-11-83 REV. O, 3-13-78 DC-1205-2.0B.I. DC-1205-2.0, REV. 2, 3-17-83, REV. O.

DC-1206-2.08, 12-13-77 REY. 2 4-28-83 DC-1206-2.0 DC-1302-2.2, REV. O. 1-6-78 REV. 5. 4-18-83 DC-1302-2.1, REV. 2, 8-28-74 ASME III. NC-2000, 16 DC-Ic04-2.0C.1, DC-1204-2.0 324.00

" I f, " " 2 REV 7, 2-11-83 REV. O, 3-11-78, DC-1205-2.08.1 DC-1205-2.0, R E 't 2, 3-17-83 REY. O.

DC 1206-2.08, 12-13-77 REV. 2, 4-28-83 DC-1206-2.0, REY. O, 1-6-78 ASME 111, ND-2000, 16 DC-1205-2.0Bl. DC-1205-2.0 325.00 CLASS 3 REV. 2, 3-17-83 REV. O, DC-1206-2.0B. 12-13-77 REV. 2, 4-28-83 DC-1206-2.0, REV. O, 1-6-78 ASTM /ASME 16 X4AQ37-6.lD. X4A037-6.lD. 326.00 COMPLIANCE, REV. II, REV. O, 3-24-80 NA/NCA--1220 12-20-84 ASME 111 NC-2000 & 16 X4AQ10-5.1 REV. X4AQ10-5.1, REV. 327.00 NE-2000 5, 12-17-84 2, 7-1-81 ASME II, l' A R T C , SFA 16 X4AQ37-5.0A, X4AQ37-5.0A, 328,00 5.1, 5.2, 5.5, 5.17 REY. 11 REV. O. 3-24-80 5.18 & 5.20 12-20-84 ASME II. PART C, SFA 16 X4AQ37-5.0A, X4AQ37-5.0A. 329,00 5.11 & S.14 REV. 11 REv. O, 3-24-80 12 20-84 O O O O O O O

i Pega No. 3 11/19/H5 IMPLEMENTATION

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MODULE 16 - SOHTED BY REFERENCE NUMBER

-=====================================

j DOCUMENT / FEATURE SECTION MODULE DESIGN LAST DESIGN FIRST CONST LAST CONST FIRST REMARKS

==================== ==-========= ====== ================ ================ ================ ================ ============= 2 ASME II. PART C. SFA 16 X4AQ37-5.0A, X4AQ37-5.0A.

5.4 & 5.9 REY. 11 REY. O. 3-24-80 12-20-84 ASME III & IX 16 X4AQ37-5.0A. X4AQ37-5.0A.

REY. 11 REY. O. 3-24-80

'12-20-84 ASME II 16 DC-1000-M-3.1.2 DC-1000-M-3.1.2 REV. 2. 4-12-78 REV. O. 5-1-73 LIMITED TO NO 16 DC-1204-4.0L. DC-1204-4.0L.

GREATER THAN 90.000 REY. 2 2-11-83 REV. 1, 3-30-79, psi YIELD STRENGTR DC-1206-4.ON. DC-1206-4.ON, REY. 2, 4-28-83 REV. 1 1-25-79 RG 1.82 6/74 16 DC-1204-1.0 DC-1204-1.0 REV. 2, 2-11-83 REY. 1 3-30-79, DC-1206-1.0 DC-1206-1.0, REY. 2 4-28-83 REY. 1 1-25-79 10CFR50 APP. A. 16 DC-1206-3.lB. DC-1206-3.IB.

GDC41 REY. 2 4-28-83, REV. O. 1-6-78, DC-1508-3.lA. DC-1508-3.lA.

REV. 6, 8-5-83 REV.6 8-5-83 .

DC-1513-3.1.lA. DC-1513-3.1.lA.

REY. O. 6-3-83 REV. O. 6-3-83 NOTE (t): REAT 16 WESTINGROUSE PROPRIETARY a EXCRANGER. ASME 111 EQUIPMENT SPECS IRFORMATION -

- SEE REMARES IMPLEMENTATIO N VERIFIED AT WESTINGROUSE TEMA-C OR TEMA-R. 16 DC-1010-72 REV. DC-1010-72, REF.

TUBULAR EXCH. MFR. 4 6-29-83 0, 8-4-78 ASSOC. CLASS C OR R h

i 4

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Poge No. 4 11/19/85 ICPLEMENTATION

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MODULE 16 - SORTED BY REFERENCE NUMBER

( ================= ====================

DOCUMENT / FEATURE SECTION MODULE DESIGN LAST DESIGN FIRST CONST LAST CONST FIRST REMARES REF NO

============ = ======-===== =====- ==-============= ================ ================ ================ ============= ==

NEMA MGl 1972 16 DC-1010-72 REY. DC-1010-T2, REV. 867.00 (NATIONAL ELECTRICAL 4, 6-29-83 0, 8-4-78 MFRS ASSOC.)

MOTORS /CENERATORS QSME Ill, CLASS 1, 16 DC-1010-T2, REV. DC-1010-72 REV. 880.00 2, 3 OR MC, NF, OR 4, 6-29-83 0, 8-4-78 CS ASME III. SUBSECTION 16 DC-1010-T2, REV. DC-1010-T2, REV. 885.00 NC, CLASS 2 4, 6-49-83 0, 8-4-78 ASME III, SUBSECTION 16 DC 'ul0-T2, REV. DC-1010-T2, REV. 888.00 NB, CLASS 1 4, J 83 0, 8-4-78 ASME 111, SUBSECTION 16 DC- 1010 -72 REV. DC-1010-T2, REV. 889.00 ND, CLASS 3 4 G-29-83 0, 8-4-78 TOTAL BREAR AREA 16 DC-1018-APP. D. DC-1018-APP. D. 963.00 LIMITED TO 12.5 SQ. REV. 2, 10-11-83 REV. 2, 10-11-83 IN.

ASME 111, NB-3650 16 DC-1018-3.38.1, DC-1018-3.3B.1, 967.00 (WCAP-8082-P-A AND REV. 2, 10-11-83 REV. 2, 10-11-83 CCAP-8172-A)

RG 1.29 REV. 3, 16 DC-1010-1.0, DC-1010-1.0, 1532.00 9/78 NEV. 4, 6-29-83 REY. 1, 4-11-79 RG 1.36 REV. O, 16 DC-1201-1.0, DC-1201-1.0, 1536.00 2/73 REY. 2, 3-9-84, REV. 2, 3-9-84, DC-1204-1.0, DC-1204-1.0, REY. 2, 2-11-83, REV. O, 3-13-78, DC-1206-1.0, DC-1206-1.0, REV. 2, 4-28-83 REV. 2, 4-28-83 RRGULATORY GUIDES R.G. 1.37, 16 X4AZol-Pl.ll.0, X4 AZO 1-Pl.ll.0, 1537.00 RRV. O, 3/73 REV. 13, 8-8-84, REV. 13, 8-8-84, X4AQOI-12.1, X4AQ01-12.1, REV. 16, 7-20-84 REY. 15, 12-5-83 O O O O O O O

Paga vn. 5 11/19/85 IMPLEMENTATION

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MODULE 16 - SORTED BY REFERENCE NUMBER

===: =====2========================e

i i

DOCUMENT / FEATURE SECTION MODULE DESIGN LJ.S T DESIGN FIR 3T CONST LAST CONST FIRST REMARES REF No

====: .e======= =========== ===-== ================ ================ ================ ================ -======e3==== =====

f ANSI N45.2.1 (1973) 16 SEE REMARES NISCO PROCEDURE NISCO PROCEDURE SRE REF. NO. 1538.00 E.S.-67 REV. C. E.S.-67, REV. A. 1537 FOR 5-31-84 9-22-82 IMPLEMENTATIO i N )

RG 1.44 HEV. O. 16 DC-1201-1.0 DC-1201-1.0 1541.00 5/73 REY. 2 3-9-84 REV. 2 3-9-84 DC-1204-1.0 DC-1204-1.0 '

REY. 2, 2-11-83 REV. O. 3-13-78 +

DC-1206-1.0 DC-1206-1.0 .

REY. 2, 4-28-83 REV. 1. 1-25-79 l RG 1.45 REY. O. 16 DC-1220-1.0 DC-1220-1.0. PROPRIETARY 1542.00 5/73 REV. 1. 9-5-80 REY. O. 5-27-77 INFORMATION -

WESTINGHOUSE IMPLEMENTATIO i " REACTOR COOLANT N VERIFIED AT PRESSURE WESTINGROUSE BOUNDARY LEAR DETECTION SYSTEM-SYSTEM i REQUIREMENTS" -  !

SEE REMARES '

BTP MEB 3-1 BTP ASB 16 DC-1018-1.0 DC-1018-1.0 1543.00

. 3-1 REv. 2 10-11-83 NEV. 0. 12-13-77 RG 1.48 REY. O. 16 DC-1201-1.0 DC-1201-1.0, 1544.00 j 5/73 REY. 2. 3-9-84 REV. O. 5-11-78 .

, RG 1.50 REV. O. 16 DC-1201-1.0 -DC-1201-1.0, 1546.00 ,

5/73 REV. 2 3-9-84 REV. 2. 3-9-84

DC-1206-1.0 DC-1206-1.0 j REV. 2, 4-28-83 REV. 2 4-28-83 X4A001-10.5 I4AQ01-10.5

, REV. 16, 7-20-84 REY. 13 11-8-82 1

3 RG 1.53 REV. O. 16 DC-1009-1.0 DC-1009-1.0 1551.00

. 6/73 REY. 2 6-3-83 REV. O. 3-23-77 1

I i i 1

i 4

i I

.~ - -_ _ _ _ _ __ _____ _ _ _ _ _-- _ ____-- _ ______ - - - _ _ - _ _ __ _ __ - _ - . .. - - _ - _ _ _ - __ _ _ _ _ . - . _ _ - - _ _ _

Page No. 6 11/19/85 IMPLEMENTATION

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e===================================

DOCUMENT /FEATUHE SECTION MODULE DESIGN LAST DESIGN FIRST CONST LAST CONST FIRST REMARKS REF NO ,

-=e ==-: - ========== ======-==-== =n==== ================ ================ ================ ================ ============= =======-

RG 1.84, REV. 20, 16 DC-1017-1.0, DC-1017-2.4.1, PROPRIETARY 1565.00 ?

11/82 REV. 3, 12-6-84, REY. O, 1-19-78, INFORMATION -

DC-1201-1.0, DC-1201-1.0, IMPLEMENTATIO REV. 2, 3-9-84 REV. 2, 3-9-84 N VERIFIED AT WESTINGROUSE WESTINGROUSE LETTER - SEE REMARKS l RG 1.85, REV. 20, 16 DC-1017-1.0, DC-1017-2.4.1, PROPRIETARY 1569.00 11/82 REY. 1, 12-6-84 REV. O, 1-19-78, INFORMATION -

DC-12al-1,0, DC-1201-1.0, IMPLEF9NTATIO REV. 7, 3-9-84, RRV. 1, 4-21-83, N VERI IED AT DC-1406-1.0, DC-1206-1.0, WESTINGHOUSE REV, 2, 4-23-83 REY. 2, 4 28-83 WES"INGHOUSE LElfER - SEE REMtRKS HEGULATORY GUIDES R.G. 1.116, 16 SEE REMARES SEE REMARES SEE REF. NO. 1578.00 REY. O, G/76 1579 FOR l IMPLEMENTATIO N

QNSI N45.2.8-(1975) 16 NISCO QA MANUAL MISCO QA MANUAL REF. FSAR 1579.00 REV. D REV. A SECT. 1.9.116 (8-10-84), VEGP (10-4-82). VEGP QAM R7, 2-85, CAM RIO, 8-72, X4 AZO 6, DIV. N1, X4 AZO 6, DIV. N1, REV. 5 REV. O, 11-21-84, DIV. 12-16-82, D1i.

N2, REV. 7 N2, REY. O, 12-17-R4 11-15-82 l RG 1.139, REV. O, 16 DC-1205-1.0, DC-1205-1.0, 1611.00 5/78 REV. 2, 3-17-83 REV. O, 12-13-77 QSME !!I, NR-3112 & 16 DC-1017-3.1.1, DC-1017-3.3.2, 1658.00 NB 3fl3 REV. 3, 12-6-84 REY. 1, 1- 24-83 O O O O O O O

I w Page No. 7 l

11/19/85 l IMPLEMENTATION

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DOCUMENT / FEATURE SECTION MODULE DESIGN LAST DESIGN FIRST CONST LAST CONST FIRST REMARES REF NO

================= ============ ====== ================ ===============.================ ====3====,= ==== ============= ====_,

i LOAD COMBINATIONS 16 DC-1017-T2, REV. DC-1017-71 REY. 1660.00 3, 12-6-84 0, 1-19-78

. ANSI N18.2-1973 16 DC-1201-2.0, DC-1201-2.0, 1727.00 kBV. 2, 3-9-84 REV. O, 5-11-78 RG 1.4 16 DC-1206-3.lB. DC-1206-3.lB. 1740.00 REY. 2, 4-28-83 REV. O, 1-6-78 RC 1.29 16 PRM PART C, DC-1010-1.0, 1754.00 SECT. 13 REV. REV. O, 8-4-78 3, 2-1-84, i DC-1010-1.0, REY. 4, 6-29-83, DMCN-2,'12-20-84 ASME III, SUBSECTION 16 DC-1201-2.0, DC-1201-2.0, 1785.00 NF REV. 2, 3-9-84 REV. O, 5-11-78 PROPAGATION OF BREAR 16 DC-1018-APP. D. DC-1018-APP. D, 1841.00 TO UNAFFECTED LOOPS REV. 2, 10-11-83 REY. 2, 10-11-83 PREVPNTED PROPAGATION OF BREAR 16 BC-1018-APP. D. DC-1018-APP. D. 1842.00-IN AFFECTED LOOP REY. 2, 10-11-83 REV. 2, 10-11-83 DOES NOT EXCEED 20%

i 0F FLOW AREA 0F i RUPTURED LINE

\

RG 1.31 REV. 3, 16 SEE REMARES SEE REF. NO. 1855.00 l 4/78 256 FOR DESIGN

, IMPLEMENTATIO i N a

- ASME III SUBSECTION ARTICLE NF 16 NOT APPLICABLE NOT APPLICABLE INTRODUCTION 1997.01 i NF 1977 EDITION WITH 1000 ONLT NO

{ wtNTER 1977 ADDENDA COMMITMENTS i

I 1

4 - _ . . ~ . . _ -. .. . .c. _. _. ...

Page No. 8 11/19/85 IMPLEMENTATION

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DOCUMENT / FEATURE SECTION MODULE DESIGN LAST DESIGN FIRST CONST LAST CONST FIRST REMARKS REF NO

================== ==-=== ===== ====== =========== ==== ================ ================ ================ ======e=====- ========

ASME III SUBSECTION ARTICLE NF 16 NISCO PROCEDURE NISCO PROCEDURE 1997.02 NF 1977 EDITION WITH 2000 E.S.-56 REV. A, E.S.-56 REV. A.

WINTER 1977 ADDENDA 6-4-84, E.S.-63, 8-27-82 REV. A, 8-22-83 E.S.-63, .' 8 V . A, 8-22-83 ASME III SUBSECTION ARTICLE NF 16 SEE REMARKS SEE REF. No. 1997.03 NF 1977 EDITION WITH 3000 880 FOR WINTER 1977 ADDENDA IMPLEMENTATIO N

ASMR III SUBSECTION ARTICLR NF 16 E.S.-8.7, REY.C,8 E.S.-8.7 REV.A,3 1997.04 NF 1977 BDITION WITH 4000 84,8.8.-100- 84,E.S.-100-WINTER 1977 ADDENDA 5,REV.C,3-9-83,E 5,REV.C,3-9-R3,

.S.-300,HEV.F 4- E.S.-300,REV.A,8 8-83 E.S.-4028-1 82,E.S.-4028

,REV.E,7-7-83,E. -1 REV.A,9-20-82 S.-4028-3,REV.F. ,E.S.-4028-3,REV l-30-85,E.S.-402 .A,9-2T-82,E.S.-

8-4,REV.F.1-30-8 4028-4,REV.A 10-5,E.S.-4028-5,RE l-82,E.S.-4028-5 V.E.1-16-85,E.S. ,REV.A.10-5-82 E

-4028-13,REV.D,8 .S.4028-13,REV.A 84,WPS-10 5-25-83,E.S.-10 20,REV.A,6-20-83 20,REV.A,6-20

,WPS-13-1-1.REV. -83 WPS-13-1-1,R A,6-20-83 EV.A,6-20-83 ASME III SUBSECTION tRTICLE NF 16 E.S.-8.7 REV. E.S.-8.7 R V.

NF 1977 EDITION WITH 5000 1997.05 C, 8-23-84, A, 3-20-84, WINTER 1977 ADDENDA E.S.-100-1, REY. E.S.-100-1, REV.

A, 8-23-82, A, 8-23-82 E.S.-100-2, REV. E.S.-100-2, REV.

E, 7-25-84, A, 8-15-82, E.S.-100-4-1 E.S.-100-4-1, REV. C. 2-1-85, REV. A. 6-21-84 E.S.-100-5, REV. E.S.-100-5, REV.

C, 3-9-83 C. 3-9-83 E.S.-Il6-1, REV. E.S.-Il6 REV.

I, 11-3-83, H, 10-12-82, F.S.-Il7, REV. E.S.-Il7, REV.

D, 6-28-82, A. 6-18-82, E.S.-Il8, REV. E.S.-Il8, REV.

C, 4-25-83 A, 9-28-82 O O O O O O O

_m.- .... .m. ..___.m u_m_... . . . _ _ - . . . . . -..._m ,~. - . , .__..-.__.m.__._.m_m.. ._.m.. . - . . ~ , . _ .

O. .

O l' a g e No. 9 11/19/85 IMPLEMENTATION

============== ,

MODULE 16 SORTED BT REFERENCE NUMBER

-============================e========

DOCUMENT /FRATUHE SECTION MODULE DESIGN LAST DESIGN FIRST CONST LAST CONST FIRST REMARES REF NO  !

==:==r=r= = = =e== ======= ====== ============-=== ================ ================ ================ ============= ==

RG l.4 REV. 2 6/74 16 WESTINGROUSE PROPRIETARY 2021.00 CALCULATION - INFORMATION -

SEE REMARES IMPLEMENTATIO N VERIFIED AT WESTIN0ROUSB COMPONENTS IN 16 DC-1205-4.0A. DC-1205-3.3B, 2030.00 CONTACT WITH BORATED REV. 2, 3-17-83 REY. O.

WATER ARE FABRICATED DC-1206-4.0A, 12-13-77 OF OR CLAD WITH REV. 2 4-28-83 DC-1206-4.0A, AUSTENITIC STAINLESS DC-1204-4.0D, REY. 1 1-25-79, STEEL OR EQUIV. REY. 2 2-11-83 DC-1204-3.3A, CORROSION-RESISTANT DC-1208-3.3C. REY. O. 3-13-78, MATERIAL. REV, 1, 3-9-83 DC-1208-3.3C.

REY. O. 5-9-78 RG 1.7 REV. 2 16 DC-1206-1.0 DC-1206-1.0 2041.00 11/78 REV. 2 4-28-83, REY. 2 4-28-83 DC-1501-1,0 DC-1501-1.0, REY. 5 REY. 5, 12-11-78, 12-11-78, DC-1513-1.0, DC-1513-1.0, REv. 0, 6-3-83, REV. O. 6-3-83, DC-1516-1.0 DC-1516-1.0, RBV. 1. 9-6-83 REY. O. 11-21-78 RG 1.61 10/73 16 DC-1005 REY. 1 DC-1005 REY. O, 2188.00 4-4-83 3-10-80, DC-1017-1.0 DC-1017-2.4.1, REV. 3, 12-6-84 REV. O, 1-19-78 RG 1.32 16 DC-1206-3.lD & DC-1206-3.lD &- 2402.00 3.1I. REV. 2, 3.lL, REV. D.

4-28-83, 1-6-78 DC-1620-3.10.13 DC-1620-3.10.13 REY. 2 3-29-83 REV. 2 3-29-83 DC-1801-1.0 DC-1801-1.0, REY. 3. 7-19-83 REY. 0, 6-24-77, DC-1821-1.0 DC-1821-1.0 REV. 5 5-2-83 -REV. 1 4-19-74

. _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ ___ __ =_ _ _ _ __ _ _ _ __ _ _ _ _ _ ___ _ _ _______ _. _ _ _ _ _ __._m.____ _ _ _ . _ _ _ . _ . _ _ _ _ _

rage No. 10 11/19/85 IMPLFMENTATION

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DOCUMENT /FE.*TURE SECTION MODULE DESIGd LAST DESIGN FIRST CONST LAST CONST FIRST REMARES REF NO

====,========== ========e=== ===e== ================ ============ === x=============== ========ee====== ============= =-==

CSS AUTOMATICALLY 16 DC-1206-3.lD. DC-1206-3.lD. 2403.00 PLACED IN OPERATION RE7. 2, 4-28-83 REV. 2 4-28-83 ON RECEIPT OF TWO OUT OF FOUR CONTAINMENT PRESSURE (HIGH-3) SIGNALS DESIGNED TO CATEGORY 16 DC-1209-7.0.1 DC-1206-7.0.1 2404.00 I AND QUALITY GROUP REV. 2 4-28-83 REV. 1 1-25-79 P REQUIREMENTS

'"E 279 16 DC-1i20-2.0 DC-1620-2.0, 2406.00 REY. 2 3-29-83 NEV. D. 7-1-77 SINGLE ACTIVE 16 DC 1206-3.10 DC-1206-3.lc. 2473.00 FAILURE PLU3 LOSP RE I. 2, 4-28-83 REV. 2 4-28-83 (INJECTION) AND SINGLE ACTIVE OR PASSIVE FAILURE PLUS LOSP (RECIRCULATION)

ASME III. CLASS 2 16 DC-1206-2.08 DC-1206-2.08.1, 2480.00 REV. 2 4-28-83 REV. 1 1-25-79 ASME 111 CLASS 3 16 DC-1206-2.08 DC-1206-2.08.2 2481.00 REY. 2 4-28-83 REY. 1 1-25-79 NEMOMGl 16 DC-1206-2.08 DC-1206-2.08.3 2483.00 REY. 2 4-28-83 REY. 1 1-25-79 MINIMUM pH OF 8.5 16 DC-1206-3.lH. DC-1206-3.lJ. 2485.00 REY. 2 4-28-83 REY. O. 1-6-78 AUSTENITIC 5.5 16 DC-1206-4.0A. DC-1206-4.04 2486.00 RET. 2, 4-28-83 REV. 1 1-25-79 RC l.1 16 DC-1206-3.lF. DC-1206-3.lF. 2512.00 REV. 2 4-28-83 REV. O, 1-6-78 O O O O O O O

O O  ! =

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MODULE 16 - SGRTED BT REFERENCE NUMBER  !

===============================,

DOCUMENT / FEATURE SECTION MODULE DESIGN LAST DESIGN FIRST CONST LAST CON 3T FIRST REMARES REF NO  ;

=== ========== ============ ====== ================ ================ ================ ================ ======t====== ==

l MAX. SPRAT PR OF- 16 0C-1206-3.15 DC-1206-3.ls, 2514.00

• 10.5 REY. 2, 4-28-83 REV. 1, 1-25-79 RG 1.82 16 DC-1206-3.lN. DC-1206-3.II, 2515.00 REY. 2, 4-28-83 REV. 0, 1-6-78

REGULATORY GUIDES R.G.-l.37, 16 SEE REMARES SEE REMARES SEE REF. No. 3201.00

? 3/73 1537 & 1538

FOR IMPLEMENTATIO N

REGULATORY GUIDES R.G.-l.37 16 SEE REMARES SEE REF. NO. 3361.00 1537 FOR IMPLEMRNTATIO N

i CVCS COMPONENTS, 16 DC-1208-3.lR. DC-1208-3.1H, 3455.00

VALVES & PIPING REV. 1, 3-9-83 REY. 0, 5-9-78 1 WHICH SEE l RADIOACTIVE SERVICE ARE DESIGNED TO LIMIT LEARAGE TO

ATMOSPHERE l

I DIFFERENTIAL 16 DC-1206-3.lN. DC-1206-3.lN, NRC QUFSTION 4153.00

PRESSURE INDUCED BY- REV. 2, 4-28-83 REY. 1, 1-25-79 0480.5 50% BLOCKAGE WITH CORRESPONDENC DEBRIS E l STARTING VOLTAGE AT 16 DC-1804-3 3F, DC-18043.3F, PROPRIETART' 4484.00

LEAST 75%'0F RATED REV. 4, 9-14-83 REY. 3, 5-25-79, INFORMATION -

! VOLTAGE OF 4 KV (AND DC-1805-3.3R, DC-18053.35 IMPLEMENTATIO ABOVE) AND 460 VOLTS REY. 4, 9-23-83 REV. 3, 7-1 N VERIFIED AT WESTINGROUSE WESTINGBOUSE i EQUIPMENT SPEC. i

- SEE REMARES N ,

E i

1 l

l 4

4 1

9 I .

Page No. 12 11/19/85 IMPLEMENTATION

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=====-=============== ====

DOCUMENT / FEATURE SECTION MODULE DESIGN LAST DESIGN FIRST CONST LAST CONST FIRST REMARKS REF NO

: =============== ============ ====== ================ ================ ===========n==== ===== ========== ============= =====

2 LOW PRESSURE PROCESS 16 DC-1208-3.38.2 DC-1208-3.3.B.2 I.E.B. 4674.00 OR HOLDUP TANKS THAT REV. 1 3-9-83 REV. O. 5-9-78 C0hRESPONDENC CAN CONTAIN DC-1203-4.0 DC-1203-4.0 RADIOACTI%E MATERIAL REV. 6 11-2-83 E C-80/08/05 REV. 5 4-15-83 ARE DESIGNED WITH FEATURES TO PRECLUDE VACUUM CONDITIONS MAXIMUM PRESSURES 16 DC-1208-1.0 DC-1208-1.0 I.E.B.

(psi) AS FLOW 4680.00 REV. 1 3-9-83 REY. O. 5-9-78 CORRESPONDENC APPROACHES ZERO DC-1: J5-1.0 DC-1205-1.0

( T t3 !.E ) .

E C-81/II/20 (AFFECTFD REV. ?. 3-17-83 REY. O.

VALVES WILL BE DC-1204-1.0, 12-13-77 MODIFIND) REY. 2, 2-11-83 DC-1204-1.0 DC- ' 206- 1. 0, REV. O. 3-18-78 RBS. 2 4-28-83 DC-1206-1.0, REV. O. 1-6-78 10CFH50 APP. B 16 SEE REMARKS SEE REF. No. 4G81.00 1537 FOR IMPLEMENTATIO N

ASME III, SUBSECTION 16 X2A005-PART I - X2AG05-PART I -

NF PRESSURIZER 4695.00 2.5.1 REV. 5, 2.5.1 REV. 4 BASE TO STEEL 4-1-82 6-29-81 SUPPORT FRAME BOLTING 2% AND 4% 16 DC-1005-T2 REV. DC-1005-72 REY.

RESPECTIVELY 4732.00 1 4-4-83 0 3-10-80 ASME B & PV CODE. SUBSBCTION 16 VEGP QAM. REV. VEGP QAM, REY. 4739.01 SECTION III. DIV. I NCA. ALL 7, 2-85 NISCO 0, 8-72 NISCO ARTICLES QAM. REV. D. QAM. REY. A.

8-10-84 X4 AZO 6 10-4-82 X4AZ06 DIV. M1, REV. 5 DIV. N1 REV. O, 11-21-84 12-16-82 X4 AZO 6 DIV. N2 X4AZ06 DIV. N2 REV. 7 REV. O.

12-17-84, GPC 11-15-82, GPC IDENT. a IDENT. &

VERIFICATION VERIFICATION MANUAL. REV. A, MANUAL. REV. 2 4-30-R4 1-15-82 9 O O O O O O

O O O O O_ ) O Page No. 13 11/19/H5 IMPLEMENTATION

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DOCUMENT / FEATURE SECTION MODULE DESIGN LAST DESIGN FIRST CONST LAST CONST FIRST REM 4RES REF NO

-- = ==========: = - ,==--===== ====== ============== = ================ ================ ================ ============= =======

ASME B & PV CODE, SUBSECTION 16 NOT APPLICABLE INTRODUCTION 4739.02 SECTION III DIV. 1 NB/NC/ND.

ARTICLE 1000 ASME B & PV CODE. SUBSECTION 16 E.S.-56 REV. D. E.S.-56, REY. A, 4739.03 SECTION III DIV. I NB/NC/ND, 6-4-84, 8-27-82, ARTICLE 2000 E.S.-6.3, REV. E.S.-63 REV. A.

A, 8-22-83 8-22-83 ASME B & PV CODE. SUBSECTION 16 NOT APPLICABLE DESIGN 4739.04 SECTION III DIV. I NB/NC/ND.

ARTICLE 3000 ASME B & PV CODE. SUBSECTION 16 E.S.-87, REV. C, R.S.-8.7, REV. 4739.05 SECTION III DIV. I NB/NC/ND, 8-23-84 A, 3-20-84 ARTICLE 4000 E.S.-56 REV. D, E.S.-56, REV. A.

6-4-84 E.S.-63, 8-27-82, REY. A, B-22-83 E.S.-63, REY. A, E.S.-300, REY. 8-22-83 F, 4-8-83, E.S.-300, REV.

E.S.-GNP-BMI-1 A, B-26-82, REV. E, 1-15-85 B.S.-GWP-BMI-1, REV. A, 6-7-83 ASME B & PV CODE, SUBSECTION 16 B.S.-8.7, REV. E.S.-8.7, REV. 4739.0C SECTION III DIV. I NB/MC/MD, C, 8-23-84 A, 3-20-84, ARTICLE 5000 E.S.-100-1, REY. E.S.-100-1, REY.

A, 8-23-82, A, 8-23-82 E.S.-100-2, REV. E.S.-100-2, REV.

E, 7-25-84, A, 8-25-82 E.S.-100-4-1 E.S.-100-4-1, REV. C, 6-21-84, REY. A. 6-21-84, E.S.-100-5, REV. E.S.-100-5, REV.

C. 3-9-83, C, 3-9-83 E.S.-Il6-1, REY. E.S.-Il6-1, REV.

I, 11-3-83 II, 10-12-82

Page No. 14 11/19/85 IMPLEMENTATION e=============

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===============e==========

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====== ====== ================ =========e====== ================ =============== ============= ==

ASME B & PV CODE. SUBSECTION 16 E.S.-4028-10 E.S.-4028-10, SECTION III DIV. I NB/NC/ND, 4739.07 REV. E, 5-21-84 REY. A. 5-23-83 ARTICLE 6000 ASME B & PV CODE. SUBSECTION 16 NOT APPLICABLE SECTION III DIV. I NB/MC/ND, OVERPRESSURE 4739.08 PROTECTION ARTICLE 7000 ASMR B & PV CODE, SUBSECTION 16 NOT APPLICABLE SECTION III DIV. I REFER TO 4739.09 NB/NC/ND.

ARTICLE 8000 SUBSBCTION NCA RCS SEISMIC 16 DC '005-APP.B. DC-1005-APP.B.

ANALYSIS-6 ret. 4-4-83 1, 4-4-83 4755.00 I, REV.

STATISTICALLY INDEPENDENT TIME -

HISTORY INPUTS:

3 TRANSLATIONAL ACCELERATIONS 3 ROTAT' '4L ACCELERAT1 xS LOADS COMBINED BY 16 DC-1017-72 REY. DC-1017-T2, REV.

QLGEGRAIC SUM 3, 12-6-84 3, 12-6-84 4757.00 DYNAMIC LOADS 16 DC-1017-T2, REV. DC-1017-72, REV.

COMBINED USING 3, 12-6-84 4759.00 3, 12-6-84 SQUARE ROOT OF SUM OF SQUARES (SRSS)

NORMAL LOADS 16 DC-1017-T2, REV. DC-1017-T2, REV.

COMBINED 3, 12-6-84 3, 12-6-84 4760.00 ALGERRAICALLY DYNAMIC LOADS COMBINED BY SRSS LOCA & SSE LOADS 16 DC-1017-T2, REV. DC-1017-72 REV.

COMBINED BT SRSS ON 3, 12-6-84 3, 12-6-84 4761.00 LOAD COMPONENT BASIS 9 9 9 9 G G G

O O O. I O.

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SUSTAINED LOADS (WT. 16 DC-1017-T2, REV. DC-1017-72, REV. 4762.00 EFFECTS) ARE 3, 12-6-84 3, 12-6-84 COMBINED WITH SRSS RESULTS BY ALGEBRAIC SUM OTHER DYNAMIC LOADS 16 DC-1017-T2, REV. DC-1017-T2, REY. 4763.00 COMBINED USING SRSS 3, 12-6-84 3, 12-6-84 RG 1.75 16 DC-1809-3.lB, DC-1809-3.lB. 4951.00 REV. 4, 1-21-82 REY. 3, 2-6-78 RG 1.75 16 DC-1809-3,lB, DC-1809-3.lB, 4952.00 REV. 4, 1-21-82 REY. 3, 2-6-78 PRIMARY MEMBRANE IS 16 WESTINGNOUSE PROPRIETARY 4987.00

< OR = 1.0 Se, EQUIPMENT INFORMATION -

PRIMARY MEMBHANE SPECIFICIATIONS IMPLEMENTATIO PLUS BENDING IS < OR - SEE REMARKS N VERIFIED AT

= 1.5 Sm WESTINGHOUSE PRIMARY MEMBRANE IS 16 WESTINGROUSE .ROPRIETARY 4991.00

< OR = 1.0 5, EQUIPMENT INFORMATION -

PRIMARY MEMPRANR SPECIFICATI0MM - IMPLEMENTATIO PLUS BENDING 15 < OR SEE REMARES N VEkIFIED AT

= 1.5 S l WESTINGHOUSE 1

RETURN LINE 16 DC-1217-3.1, DC-1217-3.1, 4996.00 AUTOMATICALLY REY. 1. 03-09-83 REV. 1. 03-09-83 ISOLATED IN CASE OF A CRACE IN RCP TRERMAL BARRIER ASME III & IX 16 SEE REMARES SEE REF. 223 4997.00 FOR CONSTRUCTION IM.'LEME NT ATIO N

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Page No. 16 11/19/85 IMPLEMENTATION

=

MODULE 16 - SORTED BY REFERENCE NUMBER

==========================

DOCUMENT / FEATURE SECTION MODULE DESIGN LAST DESIGN FIRST CONST LAST CONST FIRST REMARKS REF NO

====: =========e==== ============ ====== ================ =========_===,,= __,=_==_=_==_,__ __,__,,_=,,,,,__ _,,,__,____,_ ,,,,,,__

NO BLOCK WELDING 16 E.S.-300 REY. E.S.-300 REV. INTERPASS 4998.00 MAX 350 DEGREE F, 4-8-84 C, 8-26-82 TEMP INTERPASS TEMP E.S.-GNP-BMI-1 E.S.-GWP-BMI-1 IMPLEMENTED EXERCISE WELD PROC. REV. E, 1-14-85 REV. A, 6-7-83, THROUGH APPROVAL NISCO OAM REV. NISCO QAM REY. INDIVIDUAL D, 8-10-84 A, 10-4-8? WELD PROCEDURES FOR AUSTENITIC S/S l

1 O O O O O O O

4.0 WORK ACTIVITIES f- 1

(

4.1 DESIGN The Vogtle nuclear steam supply system (NSSS) is designed, analyzed, and manufactured / procured by Westinghouse. The primary safety.related NSSS systems designed by Westinghouse are O t. h e reactor coolant system, the engineered safeguards features system, and the engineered safeguards features actuation system. Westinghouse provides to Bechtel balance- of-plant (BOP) design criteria as dictated by the NSSS design, Westinghouse also performs the stress analysis of the reactor coolant loop and ASMK section Ill large- bore piping and the fatigue analysis

() of ASME section 111 sma ll- bore piping. In addition, Westinghouse reviews the BOP design information and drawings for compliance with Westinghouse design criteria.

Bechtel performs detailed equipment and piping layouts for the NSSS systems, and is responsible for the design and analysis of hangers, supports, restraints, etc. for piping, conduit, and components of the NSSS as described in the preceding paragraph.

Hechtel also provides information required by Westinghouse to design and supply the NSUS.

1)ivision of r.esponsibility for N stamping is defined in section  !

2.1 For those systems wit hin t he Westinghouse scope, j

. Westinghouse responsibi1ities include:

o Audits of those organizations responsible for supply of material and services associated with the installation and test of the subject systems; o Reviews of t he design specif ',ca' ions governing the piping installation, resting (including hydro), and any ,

deviations thereto; 1 J

o Reviews of piping installation, nondestructive j examination (NDE), an6 hydrotest procedures;  ;

o Witnessing of the hydrotesting of the subject systems

(~S and signing the N5 form signifying N-certificate holder

\-) approval.

4.I i 1NTERFACM CONTROL

/~N l'he design interface between Bechtel and Westinghouse for the NSSS is controlled within Bechtel by the Project Reference Manual (pHM) and within Westinghouse by internal Westinghouse procedures. Through this interface, Westinghouse provides design documents for the NSSS. Westinghouse provides a master index identifying engineering related data transmitted by C:)

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Westinghouse. The index is updated with each new data transmittal and is used as a tracking tool for transmittals.

Westinghouse data supplied by correspondence is processed within Bechtel either as correr c..dence or as supplier data depending on the scope, content, or 'n:anded use by Bechtel. This data is processed by Bechtel Drawing and Data Control (DDC) and Project Administration as descrit'ed in Appendix D.

f Bechtel also provides drawings and documents tc Westinghouse for verification o- implementation of design. Eoca discipline in responsible tot maintaining a drawing control log for the drawings to be transmitted to Westinghouse. Each disciplino it also responsible for initiating document transmittal correspondence to Westin ; house. These documents are sent by project correspondence requesting review and concurrence with a specific aspect of the design (pHM part C, section 2).

4.1.1.1 Open Items List An open items list is maintained to provide a means for assuring that actions required across the interface by either Bechtel or Westinghouse are tracked to completion. Each required interface action generated in correspondence, in vendor data which is statused as requiring revision, or in a meeting is assigned to the list and remains there until identifiable written evidence of completion is provided (pHM part C. section 2). All open items are reviewed at regular interface meetings where commitments are made to a completion schedule.

4.1.1.2 Interf ace Meetings Interface meetings between Bechtel and Westinghouse to resolve technical, schedule, er scope items are scheduled at 6 - wee k intervals and are normalty h 7!d in the Westinghouse offices.

Additional meetings are scheduled as required. Conference notes are issued in accordance with project procedures following the meeting, and assigned actions resulting from the meetings are tracked on the open items list.

4.1.2 TECHNICAL INTERFACE The following sections summarize key areas of interface between Westinghouse and Dechtel.

g 4.1.2.1 Civil /Str_uctural Di s e_i_nl i ne Bechtel developed the preliminary containment design using general arrangement dravings which depict the overall containment layout, the locations of major equipment, and the 4.1-2

routings of major piping and raceways. Westinghouse reviewed and contributed to this design.

Westinghouse is responsible for the design and analysis of the reactor coolant loop piping and e q u i pme r.t and the design and analynic of :he reactor coolant loop equipment supports. The loads on the containment structure resulting from the reactor s coolant loep equipment supports are calculated by Westinghouse

) during these analyses. These loads are provided to Bechtel for the design of the interfacing structures. Changes to the loads as the result of Westinghouse reanalysis are also transmitted to Bechtel for evaluation. Bechtel used interdiscipline criteria i g (tornado missile barriers, radiation shielding, equipment l

-s separation, etc.) and Westinghouse provided loading criteria to I a

) develop the basic design of Vogtle structural systems.

4.1.2.2 Plant Design Discipline The Plant Design Discipline is responsible for the design interface with Westinghouse for ASME Class 1 piping system analysis. Westinghouse designs and analyzes the Class 1 reactor coolant loop and pressurizer surge pipiag. Bechtel designs and Westinghouse analyzes the large- bore Cl ass 1 piping. For the small bore Class 1 piping, Bechtel designs and performs the stress analyses and Westinghouse performs the fatigue analyses. j l

'~5 Figure 4.1-1 illustrates the design interface for large-bore

,, ) Class 1 piping analyzed by Westinghouse. Figure 4.1-2  !

illustrates the control interface for small-bore Class 1 piping with stress analysis performed by Bechtel.

In addition to the pipe stress interface, drawings such as oquipment outline drawings and special Westinghouse design requirements, such as provisions for inservice inspection access and clearances f or Wes t i nghouse suppl i e:1 equipment, were transmitted to Bechtel for implementation in the plant layout pracens by Plant Design.

4.1.2.3 Mechani_ cal Discipline v

)

The Mechanical Discipline NSSS- BOP interfaces include:

o Development of process and instrumentation diagrams for Westinghouse-designed systems;

o implementation of Westin3 house process requirements for ms Bechtol-designed systems which interface with the NSSS.

l Westinghouse is responsible for the design of NSSS fluid systems. Westinghouse provides flow diagrams to Bechtel ongineering for these systems and system standard design

)

i criteria documents such as piping layout criteria. Bechtel ases 4.1 3

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these flow diagrams to develop process and instrumentation i diagrams (p& ids). From the P&lDs, Bechtel performs the same l design functions as were discussed and reviewed in Module 4 for f Bechtel- designed systems .

In the Westinghouse destor cracess, a proof- of design (POD) review is performed to confirm that Westinghouse design considerations and layou+ guidelines for these systems have been met. Bechtel provides p; ping layout and/or isometric drawings of the systems to Westinchouse for this review These drawings contain the necessary in'ormation for the Pos review (except ior Westinghouso supplied information on Westinghouse-supplied components). Westinghouce reviews these drawings for compliance with the Westinghouse denign guidelines, and also uses thece

[ drawings to compile the piping take- of f s needed for Westinghouse i

systems performance analyses. A POD review determines whether the Westinghouse-designed system as laid out by Bechtel will meet the basic functions required of that system, such as McCS flowrates versus reactor coolant system (RCS) pressure and the l residual heat removal (HHR) cooldown time of the HCS.

I Westinghouse also identifies process requirements for Bechtel- designed systems which interface with Westinghouse- designed NS3S systems. For s'fstems such a s the nuclear service cooling water system and component cooli ng wat et system. Westinghouse transmits process requirements in balance-of-plant functional requirements documents. For steam side systems such as the auxiliary feedwater system, Westinghouse process requirements are transmitted in the Steam l Systems Design Manual. These guidelines vary in detail. The l

general guidelines are included by Bechtel in the system design criteria. Specific detail requirements are incorporated into various Bechtel design documents such as calculations, control logic diagrams, and drawings.

4.t.2.4 Nuclear Discipt .-

Within Bechtel, the Nuclear Discipline has the overall responsibility for NSSS-BOP coordination.

Specific NSSS- BOP interfaces within the Nuclear Discipline include the following areas:

o LOCA and main steam line break subcompartment analysin, o Emergency core cooling analyses requiring BOP interaction; o Safety Analysis Report preparation; o Radiation shielding:

o 11a za rds ana lyses . '

4.1-4

Westinghouse performs the loss- of- coolant accident and main steam line break analyses which determine the postaccident temperature and pressure conditions inside the containment O. pressure boundary. Inputs consist of mass and energy release data from Westinghouse and containment physical dimensions, heat sink description, and containment cooling characteristics from Hechtel. The analytical results support the containmert design i criteria given by Design Criteria Document 2101, Containment Huilding.

Westinghouse provides mass and energy telease data used by Bechtel for containment subcompartment pressurization analyses.

, The results of these analyses become input to the structural design.

O Emergency core cooling analyses performad by Westinghouse determine performance specifications for the auxiliary feedwater system in the Bechtel scope. Westinghouse main steam and Ecedwater break analyses also require as input the characteristics of the HOP main steam and feedwater design.

Additionally, Westinghouse prepared information on the NSSS for the Preliminary and Final Safety Analysis Heports. Bechtel incorporated this i.nformation into the complete report.

Westinghouse and Bechtel each reviewed applicable report secttons for conformance to their scope of work.

( Hechtel performs calculations to design or evaluate shield walls enclosing radioactive equipment and materials and to predict radiation levels in all parts of the plant under various operational or accident conditions. The basic radioactive source characteristics are provided by Westinghouse in the Hadiation Analysis Manual.

Au a result of the Westinghouse fatigue analyses. WestinghTuse identifies the Class 1 pipe break locations. The break locations are used by Bechtel in performing hazard calculations.

l 4.1./.5 Electrical Discipline and Control Systems Discipline Westinghouse provides functional requirements for NSSS and interfacing systems in various functional requirement documents. Westinghouse also provides the following design documents.

o Process flow diagrams showing process instrumentation; o Functional (logic) diagrams showing system requirements such as reactor trip. T/G. trip, and initiation of engineered safety feature cystems; O

4.1 5

l o Instrumentation / control drawings such as installation drawings, rack arrangement drawings. rack interconnection drawings, and connection drawings; lh o Elementary wirinc d i v'r ams f or elect r ical- powered equipment includc r.Lipment train assignments.

Bechtel implements the Westinghouse functional requirements through various design docunents such as schematic diagrams, cabling block diagrams, control logic diagrams, and elementary h'

diagrams. The requirements are also implemeuted in the plant physical drawings.

4.1.2.6 NSSS Equip _ ment Oualification Interface Westinghouse provides safety related NSSS equipment qualified 'o generic seismic and environmental levels which are typically conservative values obtained from Westinghouse experience on other projects. The various suppliers provide Westinghouse with calculations and/or test reports which are reviewed and approved by Westinghouse to support, on a generic basis, equipment qualification at defined levels. Bechtel is responsible for locating the NSGS equipment during completion of the plant design and thereby deteruines the Vogtle specific environmental and seismic requirements.

Bechtel designers use project design criteria and Westinghouse h information documents that provide guidelines for locating equipment in acceptable environment a l conditions in their locating of NSSS equipment.

Bechtel developed programs for verifying equipment qualification levels versus location required levels for safety-related equipment. The program for comparison of environmental qualifications is ini*iated with the preparation of a System Component Evaluation Wor.: '"nat (SCEWS) for each NSSS saf ety- related component 1ccated in a harsh environment.

The operation of the evaluation is as follows:

o Bechtel identifies which components are located in the harsh environment; o Westinghouse initiates a SCEWS for each component and enters the qualification information; o Bechtel enters location environment data and compares for acceptability.

Seismic verification programs for electri. cal and mechanical NSUS equipment are similar to those for the environmental qualification verification. Westinghouse is responsible for generic' qualification o' NSSS equipment and provides to Bechtel 4.1-6

installation details consistent with their generic qualification program. This information is used by Bechtel in the design of O anchor bolts and supports to ensure that the qualified mounting is consistent with designed moanting. Deviations from Westinghouse requirements during design and installation are reviewed by Bechtel, and Westinghouse concurrence is obtained as necessary. Bechtel is then responsible for assuring that NSSS equipment is properly qualified and meets VEGP requirements.

When the qualification verification identifies an unacceptable condition, Bechtel investigates potential solutions such as relocatien, provision of special protection, or additional

, analysis. If a Bechtel solution is not found, the case is assigned an open item number for coordination between Bechtel l

O and Westinghouse. The open item is tracked on the Westinghouse open items list until resolution and closecut by documentation.

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4.1.2.7 I ns t a ll a t i o n _. Req u i r eme n t s installation requirements f or Wes tinghouse- supplied equipment are identified in various design documents transmitted to hechtel such as drawings, standards documents, or equipment specifications. These installation requirements identify the I mounting installation used by Westinghouse in the seismic and l environmental qualification of the equipment. Bechtel i

implements these requirements on drawings used for installation of t.he equipment.

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O 4.1 7

1 COMPARE CURRENT UNIT 1 PIPING /SHPPORT INFORM ATION TO AS-ANALYZED INFORMATION

• PIPING F ABRICATlON ISOMETRICS e HANGER FABRICATION ISOMITRICS

• SUPPORT STIFFNESSES

• FINAL JET IMPINGEMENT LOADS / DESIGN LOADS

• V ALVE INFORMATION AND SET UP ISOMETRIC REVISION, JET IMP.NGEMENT REVISION CONTROL FORM BPC qr E' ISSUE RE JISED YES DEVI A TIONS NO SUPPORT LOAD 2 '

ACCEPTABLE " REANALVZE '

DAT A SHE ET '

IF REQUIRED BPC E q p BPC COMPL E T E YES CWPL E TE CURRENT OPEN .S HAZARD

- ACTION $= - - --gs ACTION g EVALUATE ITEMS t TE MS BREAK NO LOCA TION IF RECutRED w AND REVISE BPC F i f IF REQUIRED REVIEW LOAD SHEE y STOP

'SSUE NEW AND SEPAR ATE TO BE REDESIGNED DS @ O SUPPORTS TO BE g REDESIGNED OR D PFE O AS DESIGNE D E

DESIGN SUPPOR T I AND ISSUE FOR I AS DESIGNED CONSTRUCTION 1P BPC ISSUE NEW PACK AGE TO W .

INF ORMATION INCLUDES-PIPING F A8RICATION AND I

ST R E SS ISOME T RICS HANGER FABRICATION DR AWING W SUPPOR I STIFFNESSES

• ISSUE UNIT 1 PIPING ISOMETRIC /

FINAL JE Y IMPlNGEMENT L CADS o V Al VE INFORMATION SUPPORT PACKAGES TO PFE FOR FCR PROCESSING J L 9P 1F RE VIE W AND COMPL E TE IF REQUIRED j DES 4GN I

W I

Fi~ AL DESIGN t

L ' '=" ' ="' =" ' - 4 REPORT LEGEND BPC BECHTEL POWER CORPORATION DC hi-R ORAWING CHANGE NOTICE - RESIDENT l

FCR FIELD CHANGE REQUEST PFL PROJECT FIELD ENGINEERING W, WESTINGHOUSE 1e 5 34 4 Piqure 4.1-1 Design Control Interfaces (Westinghouse Scope Stress Analysis) l

! _. _ __ _, ._ - _ , , _ _,. . . _ . _ . . _ , . ~ , . , _

. . - . ... ~ - - - . -

COMPARE CURRENT UNIT 1 PIPING / SUPPORT INFORMATION TO AS-ANALYZED INFORMATION PIPING F ABRICATION ISOMETRICS

• HANGER FABRICATION ISOMETRICS '

• SUPPORT STIFFNESSES FINAL JET IMPlNGEMENT LOADS / DESIGN LOADS

• VALVE INFORMATION AND SET UP ISOMETRIC REVISION, JE T lMPINGEMENT REVi$lON CONTROL FORM qy BPC BPC t

DEVIATIONS NO REANALYZE j% " "^

ACCEPT ABL E "

{ EVALUATION)

DPC J L ISSUE NEW LOAD C SHEET /DCN R TO PFE NO w

1F BPC gp PFE VES o

E ISSUE NEW SUPPORT g ,,

g LOADS AND REVIEW ACC EPT ABILITY l

l sPC i

3p BPC if W

VES UPDATE BRE AK 1

IlF REQUIRED) ISSUE NEW PACKAGE TO W DOWN LOADS AND ISSUES NEW LOADS FOR F ATIGUE ANALYSIS

'  % 7 REMAINING FIN AL =

H EVIE W PACK AGE INCLUDES: LOADS AS F ABRICATION AND STRESS DESIGNED l

ISOME TRICS WITH NODE POIN T

• THE PIPE CL ASS. SIZE, AND M ATE RI AL THE VALVE DR AWING

• THERMAL AND SEISMIC MOMENTS TP E

HEVIEW AND COMPL ETE 2 THE FINAL SUPPORT DESIGN l BPC ol w IF REQUIRED m ISOME T RIC/ SUPPORT IF REOUIRED

$g ' "

PACKAGES TO PFE FOR yl rCR PROCESSING I

I o sPC L__________4 FINAL DESIGN REPorlT LEGEND BCP 8ECHTEL POWE R CORPOR ATION DCN R DRAWING CHANGE NOTICE -RESIDENT FCR FIELD CHANGE REQUEST PFE PROJECT FIELD ENG6 NEE RING W WESTINGHOUSE 88534 4 Fir!ure 4.1-2 Design Control Interfaces (BPC Scope Stress Analysis)

i 4.2 EQUIPMENT AND MATERIALS l 1

O Safety related nuclear steam nupply system (NSSG) equipment and materials within the Westinghouse scope of supply are procured l

)

by Westinghouse in accordance with the Westinghouse Quality I Assutance Program. Technical and quality requirements, applicable to the procurement of NSSS components, are cpecified in Westinghouse procurement documents. The quality requirements

() include supplier quality assurance program requirements, requirements for access to the supplier's facility, and documentation and requirements for nonconformance control.

Generally, technical and/or quality requirements are specified

, by reference to equipment specifications and/or other documents. Westinghouse procurement documents are developed O based on input from requisitioners and are reviewed and approved by Westinghouse engineering and quality assurance. This includes review to assure that quality requirements are correctly specified, including adequate acceptance and rejection criteria; it also includes review to assure the procurement documents have been prepared, reviewed, and approved in accordance with quality assurance procedures.

The Westinghouse scope of supply is identified in Final Safety Analysta Report Table 3. 2. 2- 1 O

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0062p/323 5 1

4.3 MATERIAL CONTROL O Georgia Power Company (GPC) purchases, receives, inspects, and stores equipment and material, including welding materials required for construction at Plant Vogtle. Details of the GPC program for these activities are discussed in Appendix E. The ASME P'iler and Pressure Vessel Code, section III, division 1 requires that receipt inspection and control of ASME Code items O and material be performed by the NA installer, Nuclear Installation Services Company (N1SCO). N1SCO responsibility is limited to components and activities that are within the Westinghouse Electric Company scope of supply and are discussed in this module. Figure 4.3 1 shows material flow with references to the information contained within this text.

O 4.3.1 HEQUISIT10NING All equipment and material, including restricted consumables, are initially received and inspected at the jobsite by GPC as i explained in Appendix E. After satisfactory completion of receipt inspection, the material is placed in storage locations maintained by GPC. N1SCO obtains the equipment and material i needed for construction from GPC using the Piping Materials i Hequisition as required by procedure ES- 63.

4.3.2 RECEIPT 1NSPECTION

, Equipment and material requisitioned from GPC is receipt I

inspected by NISCO in accordance with N1SCO procedure ES-63. A I Receiving Inspection Report (R1R) (Attachment A, procedure ES-63) is used to document receiving inspection activities. The FIR is initiated by the lead engineer or decignee who inspects the equipment / material for damage and correct quantities. The HIR is then forwarded to the field QA/QC manager or designee for quality control inspection of the items. The NISCO QC inspector physically checks the items for damage, dimensions, identification, and other markings.

4.1 3 DOCUMENT HEVIEW/ ACCEPTANCE Concurrent with QC receiving inspection of equipment and matorial, the field QA/QC manager or designee reviews documentation to assure that material certifications, test reports, and Code Data reports are available and acceptable.

The review also ensures that requited documentation is attached to the RIR. All documentation is checked for traceability to the equipment and materials received and is available to the authorized nuclear inspector for review.

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After successful completion of receiving inspection and documentation review activities, the equipment or material is lh;, '

taleased for installation.

4.3.4 MATERIAL TRACEAUIb)TY Component / material traceability to supporting documentation is h maintained through installation by the following activities:

o When components or materials are issued, the RIR numbec is entered on the applicable process control sheet to

, establish traceability of the component or material to

! the location of use. h o Material is identified with code numbers and/or heat numbers, and verified by QC/QA prior to cutting the material into more than one piece.

o Components and subassemblies are assigned a unique serial number by NISCO. The serial numbet corresponds to the drawing of the component / subassembly. The serial numbers are identified on each component / subassembly, verified by QC/QA, and documented on the applicable H l it or process control sheet.

O O

O 0074p/323-5 4.3 2

.. . . - _ _ . _ m . . . . _ . _ . _ _ __ . __.

l MATE RI AL REC'VD ON SITE GPC GPC G D A -30 I REF- APP. E I i

E NG R. INITI ATES INITIATES PIPING M ATE RI AL REC'VD RECElVING INSP.

MA

~1. O TERIAL +

REQ. oBY ENGR.- - - +R E PO R T-- - y I i i l PMR ATTACHED OC REC'PT MATERIAL NISCO TO RIR INSPECT RE L E ASE D

--~ ~

OC ~

FINAL NISCO 3r DOCUMENT COMPLETeiD RIR DOCUMENT ' REVIEW TO VAULT

~~ ~-

REVIEW NOTE: ALL NISCO ACTIVITIES GOVERNED BY NISCO PROCEDURE ES43 Figure 4,3-1 NISCO Material Control f

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1 4.4 FABRICATION, INSTALLATION, INSPECTION, AND TESTING '

I 4.

4.1 INTRODUCTION

Section 4.4 describes those activities which are both pertinent and specific to the fabrication, welding, and erection of the i nuclear steam supply system (NSSS) supports, and the setting,  !

( assembly, and construction testing of NSSS cquipment, which are J within the scope of responsibilities of Nuclear Installation l Services Company (NISCO). Standard construction fabrication, 1 installation, inspection, and testing activities relative to l ASME piping, pipe supports, safety-related equipment, and

associated welding activities within the Pullman Power Products (PPP) scope of responsibilities are described in Module 4 and l Module 11.

Flowcharts in this section depict the general progression of key I

work activities, as well as key quality verification points.

Each flowchart is supplemented, as necessary, with a brief narrative which provides a more detailed description of critical activities which are specific to the particular item bei.ng l addressed.

4.4.1.1 Program Re_quirments

! l

NISCO in the prime contractor for the installation of the NSSS I

) and has an ASME approved Quality Assurance Manual (QAM) which l

complies with the ASME Boiler and Pressure Vessel (B&PV) Code section III, division 1, and 10 CFR 50, Appendix B. ASME B&PV Code section III, division 1, work is performed in accordance with the 1977 edition through the winter 1977 addenda under ASME Cortificates of Authorization numbers 2159-7 for NA installation and 2160-7 for NPT fabrication. Bechtel Power Corporation (BPC) construction specification X4 AZO 6 is the governing specification used for NSSS installation activities performed by N1SCO.

Drawings, NISCO procedures or engineering specifications, and process control sheets (PCSs) are prepared, approved, and issued for work activities in accordance with sections 2 (engineering), l'

/'N 3 (document control), 6 (process control), and 11 (control of k)

% construction processes) of the N1SCO QAM.

Drawings are provided to NISCO for installation activities by l i Georgia Power Company (GPC) Document Control. NISCO engineering {

, and QA/QC review and approve drawings prior to distribution for  ;

installation.

N1SCO procedures used for installation, fabrication, examination, inspection, purchasing, quality control, or quality assurance are assigned an engineering specification (ES) number. The ESs contain quantitative or qualitative criteria for determining that activities are satisfactorily

accomplished. The NISCO ESs are initiated by N1SCO engineering, approved by NISCO QA/QC, and statused by BPC and Westinghouse.

l Tiie principal document used by N1SCO to fabricate and install nuclear items is the Process Control Sheet (PCS', which definec specific work operations and the sequence of operations:

identifies ASME B&PV Code examinations and tests; and provideo for NISCO engineering, quality control, Authorized Nuclear Inspector (AN1), and customer signoff hold poines. When h completed, the PCS documents the process 01 Cabrication, examination and testing, and installation. rue PCS also lists the applicable drawings and NISCO engineering specifications required for a particular work activity. Figure 4.4- 1 depicts the process flow of the preparation, approval, and issuance of installation documents by NISCO.

4,4.1.2 Welding /NDE The welding activities performed at Plant Vogtle by N]SCO are i ri accordance with ASME B&PV Code section 111, 1977 edition through winter 1977 addenda. As an NA installer contracted by GPC, NISCO performs ASME B&PV Code section Ill welding of subsecLion l NF supports, pipe socket welds, and seal welds to the reactor head penetration. The welding procedures being used by NISCO at Plant Vogtle have been prepared and qualified in accordance with ASME B&PV Code sections Ill and IX and consttuction specification X4 AZO 6.

Nondestructive examinations performed by NISCO are in accordance with ASME B&PV Code sections 111 and V. Procedures being used have been reviewed by Project Engineering and Westinghouse to ensure compliance with construction specification X4AZ06 and Westinghouse requirements. Consumable materials used during the examination are of the type specified in construction specification X4 AZO 1, d ivision P- 9.

Welding and nondestructive examination (NDE) performed by N1SCO on the NSSS is depicted on Figure 4.4- 2 which presents ma jor construction activities such as:

o Transfer of material markings; o Cutting:

o Cleaning:

o Controls of welding materials; o Fit-up; o Welding; o Finish weld surface:

l 4.4-2

1 I

o NDE

o interfacing with other contractors; o Engineering QC and QA activities; l o Documentation.

'( ) Only the major work activities and principal procedures are depicted, and additional activities may be required in the l referenced procedures. l Ctattsmen perform the physical work activities under the 1 technical direction of the craft superintendent and field

,( engineers. The primary documents used by the craftsmen are the design drawings and the PCSs which specify work activities,  !

coquence, procedural references, and inspection hold points.  !

Craftsmen notify the QC inspector when a hold point is reached, at which time the inspections are performed by the inspector in accordance with and documented on the PCS.

i Once the quality control inspections are completed, the inspector notifies the ANI of any corresponding hold point.

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4.4.1.3 RiggingJ nd Handli_[tg

<~

Material and items designated for potmanent plant installation at VEGP, and which require rigging and hoisting for transporting and/or installation are classified into one of three categories. Category A items require special equipment and Project Engineering approved procedures. Category B items do j not require special equipment but do require approved '

procedures. Category C items require neither special equipment j nor specially designated handling procedures. The particular category in which an item belongs is determined by Project Kngineering and is implemented through GPC procedu re GD- T-11.

l The NSSS components for which N1SCO has direct responsibility tor rigging. hoisting, transporting, and setting include j

Category B lifts of the reactor coolant pump casings, internals,

' motor, and motor support stand. All other NSSS components (e.g., reactor pressure vessel, steam generators, etc.) were classified as Category A lifts and performed by others (i.e..

4 Ingalls and Lampson) under GpC and Westinghouse direction.

4.4.1.4 S t o r a g_e 4.4.1.4.1 Equipment Storage

, The responsibility for NSSS equipment storage rests with GPC and is controlled through the Equipment Maintenance Storage List 4.4-3

(EMSL) program which is governed by GPC procedure GD T- 09, inspection and Maintenance of items in Storage. Procedure GD-T 09 provides for the establishment of storage and f maintenance requirements for permanent plant equipment and materials at the Vogtle icbsite which conform to the parameter; of ANSI N45.2.2 (1972), ar rJpplemented by specific vendor requirements, nature of the item, and value of the item in relation to plant operations and safeguards. This is known as the EMSL program.

After receipt of an item oC equipment in accordance with GPC procedure CD A- 30 Receipt, Receipt Inspection, Storage and lla nd l ing , the mechanical discipline engineer receives a copy of.

the mechanical QC inspector's Receipt inspection Report. Usiny the criteria of procedure GD T 09, he determines whether the l item (s) should be entered into the EMSL program. For those items of equipment to be entered into the program, he initiales a Maintenance Report Card and Holodex (exhibit 01 to procedure GD-T 09) and a Maintenance Repor t-Mechanical (exhibit 05 to procedure CD T-09) which identify warehouse location, vendor requirements, vendor maintenance file number or i d e n t i t~ i e r , and the inspection interval.

The maintenance inspection requirements on the Mechanical Maintenance Report include:

o Protective covers / seals; o Desiccant / humidity; o Area / equipment cleanliness; o Compliance with Level _

(B,C D. etc.):

o Environmental / physical damage; o Preservatives / etat ;non, o Rotation performed; o Lubrication checked; o Lubrication changed; o Bearings shaf t- r.eals inspection; o lleaters energized; o Other maintenance (specify).

The mechanical discipline engineer checks the applicable inspection tequirements and enters any additional information necessary to clarify the inspection attributes for t he equi pment being evaluated.

4.4-4

The Maintenance Report Card and Rolodex is then transmitted to the mechanical QC inspection department for filing and

(-

i notification to commence maintenance inspection activities in accordance with GPC procedur e OC- T 11 Mechanical Surveillance Program. Each maintenance inspection is to be performed within an allotted interval of the actual due date as specified in procedure GD T.09.

O_,

s When an item of equipment is to be issued to the field from the GPC warehouse, the mechanical engineer algns the withdrawing contractor's requisition in the " approved by" block prior to iusue, and notifies the inspector of the new storage location I and any changes in the inspection attributes by making the necessary modifications on the Maintenance Report Card within the time allowed by procedure GD-T-09. lie also changes the EMSL file accordingly.

The rigging, handling, and transporting of equipment after issue from the GPC warehouse to the new location is performed by the requisitioning contractor in accordance with his own approved procedure. The GpC mechanical QC inspector monitors and documents these activities in accordance with GPC procedure GD-T 11, Higging, llo i s t i n g , and Transporting Permanent Plant Equipment.

rigging details See section 4.4.1.3 of this module for handling and Temporary access control barriers may be employed at the discretion of the mechanical engineer based upon the nature and O importance of the item.

Acceptable alternatives to temporary access control barriers are to limit access with lockable doors it feasible (with maintenance and inspection personnel retaining ihe keys) or to hold installation until construction is nearly complete.

Additional measures are also uned, such as fully wrapping the equipment when sand blasting, painting, or other activities are occurring in the vicinity of the equipment.

Maintenance of access control is the responsibility of GPC and all contractors on all shifts.

In controlled areas where work must be performed, the particular contractor may obtain the keys and authorization from the GPC =

equipment engineer and may unlock and watk in the controlled Cg area as required, abiding by the storaga and maintenance criteria for both the area and the included equipment.

buring internal access of equipment, tha authorized activity personnel must post a Work in Progress sign, reenergize heaters, and loosely recover equipment (if the work is not to be continued on consecutive shifts), and c.ean and properly recover /

O close equipment upon work completion.

Deficiencies found during equipment maintenance inspections and their resulting corrective actions are reported and controlled au required by GPC p r o c ed u r e GD T- 01, Nonconformance Control.

O 4.4 5

4 l

I i

t For a more detailed discussion of nonconforming item controls, refer to Appendix II.

4.4.1.4.2 Material StorIge NISCO's practice of requisitioning only the necessary amount of material at the time of installation precludes the necessity of lh NISCO maintaining material storage areas, except for storage of

weld filler material.

NISCO stores and maintains control of weld filler material in accordance with NISCO procedure ES-56, Welding Filler Material Control. ES-56 conforms to the requirements of ANSI N45.2.2 ASME B&PV Code section II, part C, and section 111, subsections h l NB and NF.

! 4.4.2 l NSSS COMPONENTS AND SUPPORTS 4.4.2.1 Reactor Pressure Vessel Supports Installation and Setting The installation of the reactor pressure vessel (RPV) supportu and setting of the RPV are depicted in Figure 4.4-3. The rigging, lifting, and transporting of the RPV was performed as a combined effort of Ingalls and Lampson (the rigging contractors) under the direction of Westinghouse and GPC. The concrete foundation pads for the RPV supports were placed by the concrete contractor, Walsh Construction Company, whose activities are discussed in Module 1.

During the setting of the RPV, the actual lifting and lowering of the RPV is performed by the rigging contractors in support of N1SCO's installation, setting, and inspection activities.

4.4.2.1.1 Reactor Pressure Vessel llead Assembly NISCO performs assembly of the head penetration appurtenances to the head at their onsite fabrication shop. The RPV head and components are received by N1SCO and, after preliminary measurements are taken, assembly commences.

As the pressure boundary work is completed on a penetration (e.g.,

attachment of the control rod drive taechanism (CRDM) and completion of welding), the penetration and apputtenance are hydrotested. Upon completion of pressure boundary work and hydrotesting of penetrations, assembly continues with installation of coil stacks, rod position indicator coils, and dummy baffle cans.

O 4.4-6

The NISCO activities which are relative to the assembly and testing of the RPV head are depicted in Figures 4.4-4 through O 4.4 7. It should be noted that, even though the ANI is not specifically depicted, he is deeply involved in review and prior notification of activities and has the option of witnessing any activities he chooses by establishing hold points on the dppllCable process sheets.

O 4.4.2.1.2 Fit- up and Assembly of Reactor Pressure Vessel Internals The reactor internals support and orient the reactor core fuel assemblies, maintain alignment between fuel assemblies and O CRDMs, absorb and transmit control rod dynamic loads to the reactor vessel flange, direct reactor coolant flow around and over the fuel elements, support the incare instrumentation, and provide gamma and neutron shielding. The reactor internals consist of the lower internals assembly and the upper internals assembly.

The lower internals assembly consists of the core barrel, lower core plate, core support, baffle assembly, secondary core support assembly, and the core support columns. The lower internals assembly is shipped assembled, with the secondary core support assembly attached, and is 33 ft high, 14 ft in diameter, and weighs 130 tons.

The upper internals assembly consists of the upper core plate, support columns, and the upper support. The upper internals assembly is 11 ft high, 14 ft in diametet across the upper uupport flange, and weighs 66 tons.

P r 'i o r to N1SCO receiving the RPV internals from GPC, it was necessary for NISCO to install the roto-lock inserts in the 11ange of the internals assemblies, to install the upper internals storage stand (U1SS) with extensions, to install the lower internals storage stand (LISS), and to assemble the reactor vessel internals lifting rig.

The U1SS and LISS are located in the refueling canal and are l

used to support the upper internals assembly and lower internals

, ,, ausembly, respectively, when these assemblies are removed from the shipping skid during construction or when they are removed from the reactor vessel. The internals are lifted from the shipping skid using the containment polar crane which is connected to the reactor vessel internals lifting rig through a O.

toad cell linkage assembly. The reactor vessel internals lifting rig is attached to either of the internal assemblies through roto-lock studs which engage mating female inserts located in the flange of the internals assemblies.

The reactor vessel and head were constructed to the 1971 through summer of 1972 addenda of the ASME B&PV Code, section III, O

4.4-7 l

division 1 which contained no requirements for Code class CS internals and core supports. Internals and core supports are designed and installed to meet the technical requirements of the h 1977 through winter of 1977 addenda of the ASME B&PV Code, section III, division 1 Jiihe"t the CS symbol stamp. The RPV internals are assembled anl .' ns t a l l ed in accordance with NISCO procedure ES- 4027- 17 and its references. The general procenc flow of the major assembling and installation activities is depicted on Figure 4.4- 8. h 4.4.2.1.3 Bottom. Mounted Instrumentation The NISCO scope of work for bottom-mounted insttumentation includes the following activities: h o Installation of the seal table; o Installation of guide tubing from the reactor vessel to the seal table; o Installation of guide tubing supports.

Currently all the guide tubes are installed with the exception of the final weld to the seal table. The seal table installation is not complete and work is in progress on the guide tube supports.

4.4.2.1.3.1 Guide Tubes. The guide tubing consists of 58 separate tubes that connect the vessel instrumentation penetration nozzles, which protrude through the bottom of the vessel, to the seal table. The preformed guide tubes are supplied by Westinghous- ir, match-marked sections that are joined during assembly by socket weld fittings. The guide tubes are seal welded to th^ real table. The seal table is a stainless steel plate apprcrinately 22 in. x 30 in, which is welded to a steel support frame. NISCO procedure numbet ES- 4 02 8-VOGTLE- 12 is used for the installation of the guide tubes.

4.4.2.1.3.2 Guide Tube Support.

O The guide tube supports consist of six separate supports, furnished by Westinghouse and installed by N1SCO using procedu re ES-4028-VOGTLE- 13. The procedure details the installation process for each support and provides details for the installation of the seal table and seal table support. l  ;

2 O l 4.4-8

4.4.2.2 Rea.ctor Coolant Pump Settina, nssembly, and Alignment

(~N The reactor coolant pumps, i nt - nals , motor support stando, and

(_) motors are rigged, lifted, transported, and set by NISCO with GPC personnel monitoring the activity. GPC supplies the cranes, 1 lifting rigs, crane operators, and transport vehicles. N1SCO l performs the rigging, and directs the lift and transportation.

1

([h s-Upon achievement of the proper pump support completion status, NISCO sets the pump casing on the supports and installs the l

l tie rods. The pump casing is then released for connection of l main loop piping by the piping contractor. Upon completion of l the piping contractor's activities, NISCO resumes assembly l operations. Figure 4. 4- 9 through 4.4-11 depicts the key '

activities in this process.

}

The pump casings, internals, motor support stands, and motors I have been installed on all four reactor coolant pumps. The installation of controlled leakage seals and coupling flanges, and the performance of final coupling alignment had not been completed at the time of assessment.

4.4.2.3 Pressurizer Installation and Setting The pressurizer is an ASME B&PV Code section 111, Class 1 vertical, cylindrical vessel with hemispherical top and bottom heads. The vessel is approximately 53 ft in height and 8 ft in

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d ia me t e r and is one of the major components in the nuclear steam supply system. The pressurizer is connected to the hot leg of one of the coolant loops by a surge line. Electrical heaters i are installed throigh the bottom head and the spray nozzle; relief and sSfety valve connections are located in the top head.

The pressurizer was designed and fabricated by the Nuclear Components Division of Westinghouse (Pensacola, Florida) in accordance with ASME BS'V Code section III 1971 edition through summer 1972 addenda.

Upon its completion in 1981, with the exception of hydrostatic r N8 testing, Westinghouse supplied the pressurizer to GPC at plant

(_) Vogtle as an NPT stamped component. After hydrotesting of the vessel during system tests, the proper Code Data Report will be prepared and the N-symbol nameplate will be attached.

The pressurizet and its associated supports are installed by

(~ N1SCO, an NA certificate holder, in accordance with ASME B&PV

(_)T Code section Ill, 1977 edition through winter 1977 eGienda.

Using procedure ES-4028 5, NISCO accomplishes the pt-cess of assembling and installing the pressurizer lateral supports an.i the final setting of the pressurizer to its supporting steel.

The installation activities performed by N1SCO are depicted on

()

v 4.4-9

Figure 4.4-12 which presents major construction activities such as:

lh o Rigging and handling; o Assembling and installing laceral supports:

o Final setting of the pressurizer lh o Interfacing with other contractors; o Engineering quality control and quality assurance activities:

o Documentation.

The major work activities and their principal procedures are depicted; however, additional activities may be required by referenced procedures.

4.4.2.4 Steam Generator Installation and Setting The four steam generators are ASME B&PV Code section 111 Clacs 1 vertical shell and U-tube evaporators with integral moisture-separating enuipment. The evaporators are approximately 68 ft in height and 15 ft in diameter and are ,

major components in the nuclear steam supply system. The steam generators are connected to the coolant loops by the inlet and outlet tube side nozzles located in the hemispherical bottom ,

head. I l

The steam generators were designed and fabricated by the Nuclear Components Division of Westinghouse (Pensacola, Florida) in  ;

accordance with ASME B&PV Code section Ill, 1971 edition through  !

summer 1972 addenda.

Upon its completion, in 1981, with the exception of the hydrostatic testing, Westinghouse supplied the steam generators  !

j tc CPC at Plant Vogtle as an NDT stamped component. After '

h) x

'esting of the vessel during system tests, the proper Code Dac Report will be prepared and the N-symbol nameplate will be attached.

The steam generators and their associated supports are installed by N1SCO, in accordance with ASME B&PV Code section III, 1977 edition through winter 1977 addenda. Using procedure ES- 4028- 3, NISCO accomplishes the process of assembling and installing the steam generator vertical column assemblies and upper and lower lateral supports, as well as the final setting of the steam generators.

O 4.4-10

I The installation activities performed by NISCO are depicted on Figure 4.4-13 which presents major construction activities such

o Rigging and handling

o Assembling and installing supports:

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4.5 TURNOVER TO GPC

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5 Upon completion of Nuclear Installation Serviceu Company (NISCO) field work on an item, the NISCO field OA/QC manager ensures that the applicable process control sheets and the supplemental quality documentaton and records are collected, reviewed, and compiled into a package in accordance with N1SCO procedure ES 126. Documentation for work activit es which do not require O

an ASME data report and certification by the authorized nuclear inspector (ANI) or code stamping, is then transmitted to Georgia (GPC), where receipt is acknowledged by signature on the NISCO transmittal.

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l

'> . 0 AUDITS  ;

igI This module section contains a discussion of the Quality

\s' Assurance (QA) audit process, Nuclear llegulatory Commission i (NHC) inspections, and special evaluations addressing pipe 1

treus analysis and pipe supportu. Throughout the Plant Vogtle l construction program, onsite audit.s have been performed oy Georgia Power Company (GPC) QA. In addition to regularly (3) scheduled and periodic inspections and i.nve s t i g a t i o n s , the NHC

'" nas conducted a Systematic Assessment of Licensee Performance ovaluation and a special investigation performed by the Itegional Construction Ass m sment Team. Plant Vagtle was the first of 22 g

atility siter that initially participated in onsite )

, t nves t iga t. i ons and evaluat. ions performeu by the Institute of Nuclear Power Operation (INPO). An off-shoot of the pilot INPO (V) l program and the subsequent follow up onsite investigation was ihe format. ion of the GPC Self-Initiated Evaluation program.  ;

1 iho contont of t.his section is divided into four s u bs e c t. i o ns :

Seetion Tit 1e l

'>.1 Project Organization Audits

'; . / NRC Inspections b 4 Past Construction and Design Problems

'.4 Gupplementa1 Aud i t.s f

(3 U 't he Headiness Heview discipline teams uaed the results of the e.irious audits, evaluations, programs, inspections, and records of p a s t. problems as an aid in developing a meaningful and ,

r op r esiin t a t ive assessment program. l l

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5.1 PROJECT ORGANIZATION AUDITS r

k_ 5.1.1 AUDITS OF DESIGN ACTIVITIES 5.1.1.1 Georgia Power Company Audits f)

'V This section addresses the Georgia Power Company (GPC) audit findings applicable to Module 16. The Readiness Review Team has identified eight GPC quality assurance (QA) audits addressing subjects within the scope of Module 16. These audits resulted in the identification of six findings. As part of the Readiness Review program, the findings were reviewed and categorized by subject in the audit matrixes.

(

The QA evaluation cf the audit findings included a determination of impact on safety-related work, if any, corrective measures, and assessment of the proposed actions to minimize the potential for recurrence. Significant audit findings which addressed nuclear steam supply system (NSSS) design interface control is discussed below in more detail.

o AFR 468 Six Westinghouse drawings, which had been superseded were found issued for use, and seven Westinghouse fg letters were not entered in the records management

( system (RMS). The project supplier document register was reviewed against the Westinghouse Master Index to assure that voided and superseded Westinghouse documents have been removed from the jobsite active file. Drawing transmittal letters had been erroneously interpreted as not required in the RMS. All post transmittal letters were ent'ered in the RMS as corrective actien. The Project Reference Manual was revised for processing of voided / superseded Westinghouse documents and project personnel were trained in the revised procedure.

o AFR 533

(~s A review of installation procedures in use by the q contractor, Nuclear Installation Services Company (NISCO), found several instances of conflicts or inconsistencies with the Bechtel Power Corporation (BPC) installation specification and also conflicts with basic Westinghouse guidelines or requirements. Westinghouse -

Bechtel - NISCO coordination led to resolution and O changes in NISCO procedures.

1 o _AF R 573 Page 1 of 2 was missing from erection contractor's welding procedure specification. Bechtel had failed to deliver the page at the time ~of transmittal and omission was not detected by NISCO review. Procedures for reviewing documents for adequacy were reexamined and individuals involved were reinstructed in their use.

5.1.1.2 South _ern Comp _any Services Audits l

Southern Company Services (SCS) audits BPC to ensure adequate )

implementation of the interface relationship between BPC and Westinghouse as N-certificate holder of record for the NSSS.

Two findings within the scope of Module 16 were identified. In both, SCS questioned the Westinghouse position that BPC design documents do not require Westinghouse review. The Project Reference Manual was revised to clarify the BPC/ Westinghouse relationships concerning design documents.

5.1.1.3 Other Desian Audits in addition to the audits of Bechtel, both Bechtel and GPC have performed audits of Westinghouse design activities. From July 1983 through June 1985 Bechtel, in conjunction with GPC and/or SCS, performed 12 audits of Westinghouse compliance with the Westinghouse QA program. No significant findings were identified as a result of those audits.

The results of the Bechtel/GPC audits of Westinghouse complied with the frequency review of Westinghouse by the Nuclear Regulatory Commission and other customers, lead the Readiness Review Team to conclude that the Westinghouse design program has adequately adhered to the Westinghouse QA program.

5.1.2 AUDITS OF CONSTRUCTION ACTIVITIES GPC conducts regularly scheduled and planned audits of g construction activities to verify compliance with project W commitments and suitable response to procedural and design requirements. Audit findings and recommended corrective measures are reported to the management of the audited organization. The audit results are also reported to GPC management and the GPC QA organization tracks the acceptability and timeliness of response. The QA organization issues monthly reports to management summarizing audit results and response status.

l l

5.1-2

This section addresses the GPC audit findings applicable to Module 16. The Readiness Review Team has identified 30 GPC QA audits addressing subjects within the scope of Module 16. The subject categories of each finding are presented in the audit matrix at the end of this section. The 30 audits resulted in the identification of 22 findings and observations related to construction activities during the period from March 1979 through May 1985. As part of the Readiness Review program, the

() findings were reviewed and categorized by subject in the audit matrixes. The results are summarized in the table below:

Subiect Area Frequency (a) Findinq(b)

Materials 15 6 Training and qualification 2 0

() Fabrication Inspection 11 9

5 3

Testing 4 O Measuring and test equipment 4 1 Document control 7 3 QA records _8 i TOTALS 60 22 The QA evaluation of the audit findings included a determination of impact on safety-related work, if any, corrective measures, and assessment of the proposed actions to minimize the potential for recurrence. As of May 31, 1985, four of the findings N remained "open":

\

Audit No. Finding No. Date Description GD07-83/64 468 7/14/83 Control of Westinghouse documents MD07-84/44 652 6/14/84 NISCO procedures not responsive to Westinghouse requirements GD08 85/11 759 3/22/85 Improper NISCO welding procedures O MD09-85/10 765 02/28/85 Contractor responsible for cleaning of external surfaces of fluid systems does not have an approved QA manual Resolution of these items is in progress.

a. The Frequency column includes findings, observations, and other references to the subject areas within audit reports.

b. The Findings column includes findings, observations, and O deficiencies and reflects multiple listings of the same finding if more than one category is affected.

5.1-3

Audit findings which addressed the subject categories of materials, QA records, and fabrication are discussed below in more detail.

5.1.2.1 Materials Two of the findings required Westinghouse participation in developing a response or corrective measures. Both were h isolated instances and were not indicative of programmatic problems.

o 052-0BS Clarification was required to establish when GPC assumed O responsibility for surveillance of storage conditions of

NSSS equipment.

o AFR 493 The Westinghouse-furnished covers for the HPV lower internals did not allow for a complete seal as required by Westinghouse documents. The covers were evaluated by Westinghouse and found to be acceptable.

i The remaining findings involved improper preparation and disposition of Discrepancy Reports on NSSS bolting and improper {

storage and inspecti3n of RPV internals. <

All findings were satisfactorily closed.

5.1.2.2 Fabrication Three of the findings in the fabrication category were summarized in section 5.1.2 above and are still open. AFR-409 involved N1SCO's performance of work without the correct process control sheet, and AFR-533 addressed a N1SCO welding procedure which was not fully responsive to Westinghouse requirements.

5.1.2.3 OA Record Of the four findinge in this category, one (AFR 409) involved activities performed on the steam generator without a process control sheet to document the work. The others were examples of incomplete or inaccurate entries or documents.

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Q O Page No. I 11/14/85 DESIGN AUDITS

=

MODULE 16

=

EDIT INIT AUDIT DATE MOD DESIGN CALCUL DRAW- SPEC NO ORGAN SUPPLR 'DEVIAT. TRAIN DERIGN DESIGN DESIGN MISC WEST.

NUMBER CRIT ATIONS IN05 BATA REPORTS PROM REVIEW DOC CNT CEANGES SCOPE

========== ========== ======== === ======= ======= ======= ======= ======= ======= ======= ======= ======= ======= ======= ===

883 BPC-QA 06-01-78 16 X

884 BPC-QA 02-01-79 16 I 885 BPC-QA 12-01-79 16 X

888 SPC-QA EG-8 07-26-84 16 g 890 BPC-04 VH-lli-13 06-16-85 16 x r 886 BPC-QA VB-Ill-9 11-24-81 16 I 3 218 889 BPC-QA VH-Ill-9 08-02-84 16 3 3 77 BPC-QA VN-III-9 09-30-80 213 g I

16 79 BPC-QA VM-III-9 07-01-82 16, 3 x 218 80 BPC-GA VR-III-9 07-19-83 16 N X 219 188 GPC-QA 0807-83/64 07-22-83 218 468 16 469 GPC-0A MD07-82/II 09-08-82 16 I 2

219 GPC-QA MD07-83/10 12-12-83 16 533 633 9

220 GPC-QA MD07-83/18 04-14-93 16 408 472 GPC-QA N307-83/73 08-3-83 16 I 221 GPC-04 M007-84/II 02-27-84 16, 573 21G 21 F

Pete No. 2 11/14/85 DESIGN AUDIT 8

=

MODULE 16

======

EDIT INIT AUDIT DATE MOD DESIGN CALCUL DRAW- SPEC NO ORGAN NUMBER SUPPLR DEVIAT. TRAIN DESIGN DB5IGN DE5 ION MISC WEST.

CRIT ATIONS INGS DATA REPORTS PRGM

=== ========== ========== ======== === ======= ======= ===,,s

REVIEW DOC CNT CHANGES SCOPE

===.3== .... = ======= ======, =.... =...... ..==,== 3...... .==....,

I kl8 GPC-QA MD07-84/44 07-09-84 16 654 652.655 580 GPC-QA SP01-85/30 04-29-85 16 I 21F 21 0

714 NRC-INS 82-24 11-03-82 18 E

270 5C5-GA N/A 11-12-84 213 84-17,8 E 21 84-15 4-19 J,1 6

l e e e G 9 9 9

O O O O 0 C 3 CONSTRUCTION AUDITS EDIT lttlTIATING AUDIT MATE- TRAIN / FABRl- INSPEC- TESI- MEASURE DOCUMENT QA NO. ORGANIZATION NUf0ER DATE MODULE RIAL QUAL CATION TION IPlG & TEST EQ CONTROL RECORDS REMRKS EXPLANATION OF FIELDS EDIT NO. Internal reference numbers INillATING ORGANIZATION lhe organization performing audit or inspection:

GPC - QA =. Georgia Power Company QA Department HART-N-616 = Hartford Steam Doller and inspection Conpany NISCO e kuclear Installation Service Company NRC-lNS = Nuclear Regulatory Cornission inspection Report West = Westinghouse BPC r Bechtel Power Corporation SCS e Southern Conpany Services I NPO = Institute of Nuclear Power Oper ations AUDil NUiRR - i dea l i f i ca t i u.. i.undan of audit or ins .: tion assig .cJ by initiat ing organiratim DATE - Date of audit or report receipt date MODULE Readiness Review module number MATERIAL - Material, storage, damage, handling, cleanliness, etc.

TRAIN / QUAL - Training and qualification of personnel FABRICAllON - Manufacturing / installation activities INSPECTION - Inspection and nondestructive examination TESTING - Pressure tests, flow tests, load tests, etc.

MEASURE & TEST EQ - Measurement and test equipment DOCUENT CONTROL - Document control l

QA RECORDS - Quality Assurance records 0009p/325-5/3

O rgge No. 1 O O O O O O I171H/85 CONSTRUCTION AUDITS

======

MODULR 16

======E

EDIT INITIATING AUDIT DATE MODULE MATERIAL TRAIN / FABRIC- INSPECT- TESTING MBASURE DOCUMENT QA REMARE5 NO ORGANIZATION NUMBER QUAL ATION ION & TEST RO CONTROL RECORDS

3= ====-====================e= ======== ====== ======== =e===== ======= ======== ======= ======== ======== ======= ==========e=

327 GPC-QA GD07-83/64 07-14-83 16 468 332 GPC-QA 0007-84/27 04-10-84 16 613 380 GPC-OA 0D08-85/11 03-22-85 16 759 451 GPC-QA MD03-82/85 08-03-82 16 320 495 GPC-04 MD05-81/31 05-19-81 16 I I 517 GPC-QA MD06-82/101 11-29-82 16 I I R

$20 GPC-oA MD06-83/23 04-27-83 16 I I I 526 GPC-oA MD06-84/76 11-07-84 16 I I I I I 530 GPC-QA MD07-82/142 12-21-82 16 052-085 379 .

379 532 GPC-QA MD07-82/66 06-15-82 16 302 534 GPC-QA MD07-83/109 12-12-83 16 533 X X 536 GPC-QA MD07-83/IB 04-14-83 16 409 409 537 GPC-QA MD07-83/43 06-30-83 16 441 539 GPC-04 MD07-83/76 08-24-83 16 493 1422 GPC-QA MD07-84/Il 02-06-84 16 X X 574 574 544 GPC-QA MD07-84/44 07-09-84 16 652 653 652 578 GPC-CA MD09-85/10 02-28-85 16 765 579 GPC-04 MD10-82/134 01-26-83 16 I

$82 GPC-04 MD10-82/38 03-31-82 16 I I I

Page No. 2 11/1H/85 CONSTRUCTION AUDITS

=======

MODULE 16

=

RDIT INITIATING AUDIT DATE MODULE MATERIAL TRAIN / FABRIC- INSPECT- TESTING MEASURE DOCUMENT QA REMARIS NO ORGANIZATION NUMBER QUAL ATION 10N & TEST IQ CONTROL RECORDS

== === ======== =============== ======== ====== ====re== ======= ======- ======== ======= ======== ======== ======= ===========

583 GPC-0A MD10-83/45 06-03-83 IS I I I I I 584 GPC-OA MD10-84/01 02-13-84 16 I 633 GPC-QA MD13-79/07 04-11-84 16 I I I 635 GPC-QA MD13-80/19 05-13-80 16 I I 638 GPC-QA MD13-80/44 12-10-80 16 I I 640 GPC-QA MD13-81/06 02-13-81 16 I 1431 GPC-QA MD13-84/51 07-10-84 16 I 667 678 GPC-QA MD14-84/65 09-25-84 16 I I 773 GPC-QA WH01-82/47 04-15-82 16 291 291

• 784 HART-N-626 01 07-05-84 16 I I I I 1.0-DEF. 2.0-DEF ANSI-N-626.0 2.0-DEF AUDif 785 HART-N-626 C2 03-04-85 16 I I ANSI-N-626.0 AUDIT 787 NISCO 4027-Al 01-05-83 16 I I SUPPLIER AUDIT 788 NISCO 4027-A10 01-24-84 16 I I I SUPPLIER AUDIT 789 NISCO 4027-All 01-10-84 16 I I SUPPLIER AUDIT 790 NISCO 4027-Al2 03-06-84 16 I I I SUPPLIER AUDIT

79) MISCO 4027-A13 04-24-84 16 1.0 16.0-095 7.0, 3.0, 2.0 20.0-055 13.0 11.0 . 22.0 19.0 23.0-085 S,0 0 0 0 0 0 0 0

e m_ _4- _ . _ _ - _ . . . . . _ . _ . _ _ . _ _ _

hae No. 3 It*IH M5 CONSTRUCTION AUDITS

==-=~======-==-

MODULE 16

=

EDIT INITIATING A TID I T DATE MODULE MATERIAL TRAIN / FABRIC- INSPECT- TESTING MEASURE DOCUMENT QA REMARKS NO ORGANIZATION NUMBER QUAL ATION ION & TEST EQ CONTROL RECORDS

-- =e==--===== -========-===== === ==== ====== ======== e====== ======= ======== ======= ======== ====== ======= ==== ==== ==

792 NISCO 4027-A14 08-16-84 16 I I I I 793 NISCO 4027-A15 06-06-84 16 11-095 I ,

1 794 NISCO 4027-A16 07-10-84 16 I I 795 NISCO 4027-A17 11-08-84 16 8.A 12.0 1.0-083, 2.0-088 7.0 .

8.A. 3.0-085 8.8 8.C . 3.B.

7.0, 8.A.

11.0 797 NISCO 4027-Al9 02-18-85 16 I 10-1.10 I 35.0 I 4.0 3 -2 798 NISCO 4027-A2 02-16-83 16 I I SOURCE INSPECTION I

799 NISCO 4027-A3 04-22-83 16 2.0 7 0-088 7.0-085 801 NISCO 4027-A4 04-26-83 16 I I I 10.5-0B5 I SUPPLIER SURVEY 802 NISCO 4027-A5 06-28-83 16 I t

J 803 NISCO 4027-A6 07-28-83 16 I 804 NISCO 4027-A7 10-06-83 16 I I 805 NISCO 4027-A8 12-12-83 16 I I I 3.0-088 f

I 806 NISCO 4027-A9 12-14-83 16 I 807 NISCO 4027-CAR-01 05-31-84 16 I I I RESPONSE TO N-626 AUDIT

, 809 NISCO 4027-51-1 11-02-82 16 I SURVEY OF MACNINING 3 SUB-CONTRACT OR

}-

j i

) '

t

Pnge No. 4 II/lR/85 CONSTRUCT 100 AUDITS

============

MODULE 16

=

DATE MODULE MATERIAL TRAIN / FABRIC- INSPECT- TESTINO MEASURE DOCUMENT QA REHARES FDIT INITIATING AUDIT NO ORGANIZATION NUMBER QUAL ATION ION & TEST EQ CONTROL RECORDS

- =- :====== -=== =============== ======== ====== ======== ======= ======= ======== ======= ======== ==.===== ======= ============

4027-51-3 02-09-84 16 SUPPLIER 1 811 NISCO I I AUDIT 812 NISCO 4027-SI-4 06-28-84 18 I I SUPPLIER AUDl?

813 NISCO 4027-5I-5 07-26-84 16 I I SUPPLIER AUDIT 814 NISCO 4027-51-6 09-05-84 16 I I SUPPLIER AUDIT BJ8 NRC-!NS 77-03 07-06-/7 16 #7-uJ-M1 I 17-03 N1 905 NRC-INS 79-14 09-11-79 16 I I 921 NRC-INS 80-01 02-21-80 16 I 941 NRC-INS 80-10 07-03-80 16 80-10-03 I I 960 NRC-INS 80-15 10-28-80 16 80-15-02 983 NRC-INS 81-07 07-27-81 16 81-07-01 81-07-02 996 NRC-INS 81-11 12-02-81 16 I 100! NRC-INS 81-13 12-28-81 16 I 1009 NRC INS 82-02 02-02-82 16 I 1015 NRC-INS 82-04 02-23-82 16 I I 82-04-0 1

1019 MRC-INS 82-05 03-31-82 16 I 1025 NRC-INS 82-06 04-12-82 16 82-06-02 I I I O O O O O O O

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5.2 NUCLEAR REGULATORY COMMISSION INSPECTIONS l

5.2.1 NRC INSPECTIONS DESIGN l

The nuclear steam supply system (NSSS) design has been addressed in 12 Nuclear Regulatory Commission (NRC) inspections. During these inspections no violations related to the NSSS were O identified. The design audit matrix identifies the specific NRC inspections with the subject areas evaluated identified by an X.

b.2.2 NRC INSPECTIONS CONSTRUCTION NSSS equipment, materials, or construction activities have been 4

addressed in 35 NRC inspections during the period from July 1977 through April 1985. During these inspections, the NRC identified 6 violations, of which only the most recent, 85-09-01, addressed inadvertent contact of the stainless steel cladding of the reactor pressure vessel by a carbon steel personnel basket, remains open. For further information on the open item see NRC Inspection Report 85-09. The NRC Inspection Heports addressing the construction activities included in the J

scope of Module 16 were reviewed and categorized into one or more of eight subject areas as shown below:

Frequency Number

'( Subject of of Areas Inspection Violations (d)

Materials 21 4 Training and Qualification 1 0 Fabrication 17 0 Inspection 9 .1 Testing 4 1 Measuring and Test Equipment 0 0 Document Control 0 0 QA Records _6 0 58 6 The audit matrix at the end of this section identifies the specific subject categories assigned to each of the violations, O unresolved items, inspector followup items, and licensee-identified deficiencies. For convenience, each of the I

violations has been circled for toady identification (subject areas which were evaluated but did not result in violations, unresolved items or inspector follow up items are identified by {

The violations were not repetitive in nature, indicating O an X). '

that the few procedural violations identified were isolated a [ ~ NI5~l'Et i o n' 85 09-01 has been listed in two subject areas resulting in a total of six.

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l occurrences. Violation 83-18-01, though identified at the VEGP construction site, addresses improper or incomplete actions by the reactor pressure vessel manufacturer.

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5.3 PAST CONSTRUCTION AND DESIGN PROBLEMS 5.3.1 DESIGN No significant design problems were uncovered during the course of Georgia Power Company (GPC) and Becht el Power Corporation (BPC) reviews and the Institute of Nuclear Power Operation O- (INPO) pilot evaluation program. However, six significant design problems were identified during evaluations by Westinghouse. In all cases the Nuclear Regulatory Commission (NRC) was notified of a reportable, or potentially reportable,

, deficiency.

5.3.1.1 Gate Valve Closure Problems Gn October 29, 1980, Westinghouse advisad GPC that during pr eopera t iona l testing on another domestic plant, problems were '

encountered on 3- in. gate valves supplied by Westinghouse Electro Mechanical Division. The valves did not accomplish full stroke during testing. As a result of ;his problem, GPC requested BPC and Westinghouse to determine applicability and evaluate the effect of similar 3 in. valves on Vogtle.

As a result of the evaluation, it was determined that four of the subject valves were to be used in safety-related piping nystems. It was concluded that failure of the valves to close

('N

(_) constituted a reportable deficiency and a reportable safety hazard. To resolve the valve problem, the valve actuators were modified to ensure closing. The corrective action was reported to the NRC on April 21, 1981.

I h.3.1.2 Gate Valve Closure Problems l>uring the evaluation of the 3-in. yate valves, Westinghouse determined that a similar problem existed for cer tain 4- in. yate valves. It was determined that failure of t he 4- i n. valves to close also constituted a reportable deficiency and corrective action to modify the valve actuators was reported to the NRC on July 2, 1981.

5.3.1.3 Sheared Pinion Keys in_ Valve Motor Operators On August 4, 1982, Westinghouse reported to the NRC and GPC that sheared pinion keys had been discovered in six 1,imitorque Model O- GB 0 25 valve motor operators. A subsequent evaluation by BPC revealed that this particular Limitorque nodel would not be used at VEGP; however, Westinghouse found tha' several other Iimitorque models (J B- 0. SB-00, SMB.0, a nd Sb'U- 00) used keys made I from the same material, Series 1113 resulft.tzed carbon steel.

C:)

Since GPC discovered there are approximately 50 valves with the above model Limitorque motor operators at VEGP, they reported to the NRC that both a significant deficiency and a substantial h sat ety hazard exist.

To eliminate potential fa! Lute, Westinghouse and Bechtel will furnish replacement pinion keys of certified AIS1 1018 material which VEGP is to install during future maintenance cycles for h valves with the applicable model numbers. Georgia Power Company reported this corrective action to the NRC on June 1, 1983.

5.3.1.4 Charqi_n_q, Pumps Modifications, implemented as a result of 1&E Bulletin 79 06A. O resulted in 1 condition in which the charging pumps could operate at shutoff pressure for an extended period in the event of secondary side high energy line break or spurious safety injection signal. Westinghouse identified this problem to Gpe on May 8, 1980. This event could lead to degradation of the charging pumps and tesult in the pumps' inability to deliver required design flow. Corrective action to provide an auxiliary miniflow path for each charging pump was reported to the NRC on August 10, 1982.

5.3.1.5 Ungualified I,imitorque Valve _ Operator Westinghouse was informed by another domestic power plant that motors on four Limitorque operators for valves in the charging pump auxiliary miniflow lines were not qualified to appropriate IEEE standards. Westinghouse advised GPC on October 5, 1984 that this problem also applied to valves supplied to VEGP.

Since the motors were not qualified, the valves could not be assumed to operate as required, and this condition could cause damage to the chargin7 rumps. It was determined that this condition represented a reportable deficiency, and correctivo action to replace the operators with qualified operators was reported to the NRC on December 19, 1984.

b.3.1.6 Westinghouse Gate Va.lve Position Indication l

On December 23, 1982 Westinghouse advised GPC of a potential ,

safety concern involving valve position indication for certain i Westinghouse manufactured gate valves. The valves had been wired in a manner such that the position indicator would indicate a valve was fully closed prior to the valve disc fully isolating flow. If the valve were to stall or bind after the {

premature closure indication, the plant operator would have an i inaccurate indication of the valve position, and the potential would exist for the unit to be in an unanalyzed condition.

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j lt was concluded that this condition represented a reportable  ;

concern and corrective action to rewire the valves to provide l

} accurate position indication was reported to the NHC on l l November 9, 1983.

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') .3.2 CONSTRUCTION i

1 There have been no significant past construction problems identified.

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5.4 _ SUPPLEMENTAL AUDITS l

I The ASME Boiler and Pressure Vessel Code, section 111,

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subsection NA/NCA addresses the need for quality assurance programs and imposes certain audit activities on specific organizations. This section addresses those audits applicable to this module.

O The ASME surveys N-type certificate holders triennially for continuation of authorization to extend their ASME certificate for work performed at the VEGP construction site. Data from these surveys are confidential and unavailable for inclusion in this module.

O 5.4.1 AUTilOR17,ED NUCLEAR INSPECTION AGENCY AUDITS ANSI N-626 requires biannual audits of certificate holders. The two audits of Nuclear Installation Services Company (NISCO) activities identified procedures problems related to Authorized Nuclear Inspector acceptance of quality assurance manual revisions and review of Discrepancy Reports. No programmatic problems were noted. The Authorized Nuclear Inspection Agency at VEGP is the 11ar tf ord Steam Boiler Inspection and Insurance Company.

5.4.2 WESTINGilOUSE AUDITS Westinghouse is the responsible N- certificate holder for the nuclear steam supply system (NSSS) primary loop piping and portions of other NSSS systems. As the responsible N certificate holder, Westinghouse has the responsibility for j auditing GPC activities as the material supplier for contractors l performing work within the Westinghouse scope of contract j reaponsibilities, i

Westinghouse has conducted four new audits and one followup audit during the period from January 1982 through May 1985 in which they identified five findings and nine observations. The Headiness Review Team has categorized these findings and

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% observations into one or more of nine subject categories as

\_ summarized in the table below.

_S ubj ec t A r ea F r eq u e n,cy ( a ) Findings (b)

Design NA NA O Materials Training and qualification 8

1 0 5

Fabrication NA NA Inspection 5 2 Testing NA NA Measuring and test equipment NA NA O

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j Subiect Area Frequency (a) Finding _s(b) l Document control 7 6 QA records 9 7 l TOTALS 30 20 l

The findings were related to approval of the GPC Identification and Verification Manual, verification that suppliers of welding materials were acceptable to Westinghouse anc appeared on the Westinghouse list of qualified suppliers, and the use and control of purchasing documents. The observations addressed numerous subjects and included recommendations and comments concetning GPC activities and QA program requirements but no programmatic problems were identified.

All findings have been satisfactorily closed.

5.4.3 N1SCO INTERNAL AUDITS The N1SCO internal audits were frequent, effective, and identified numerous procedural and QA program deviations. The timeliness of the audits and the rapid remedial and corrective l'

actions assured that none of the findings resulted in significant problems.

As in previous sections, the findings and observations have been listed on matrixes and categorized into specific subject areas as tabulated below.

Subiect Area Frequency (a) Findingg(b)

Design (not applicable) 0 0 Materials 9 5 Training and qualification 4 2 Fabrication 5 1 Inspection 6 4 Testing 0 0 Measuring and test equipment 2 0 Document control 15 10 QA records 17 13 TOTALS 58 35

'a . The Frequency column includes findings, observations, and O

i other references to the subject areas within the audit i reports.

b. The Findings column includes findings, observations, and deficiencies, and reflects multiple listings of the same item if more than one category is affected. l 0025p/323-5 \

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6.0 PROGRAM VERIFICATION

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. (_ f This section describes the activities undertaken to ascertain whether the design and construction work processes have been  ;

adequately controlled to ensure implementation of licensing commitments and that the results of these work processes conform to project procedures and design requirements.

!() This section is divided into two parts. Section 6.1 describes activities related to design program verification, and section 6.2 addresses construction program verification.

. e Resulting findings have been classified intc levels of importance to plant safety. The following levels are used:

1 - Violation of licensing commitments, project procedures, or engineering requirements with 4

indication of safety concerns.

11 - Violation of licensing commitments or engineering requirements with no safety concerns.

111 - Violation of project procedures with no safety )

concerns.

Tables 6.1-2 and 6.2-1 provide a listing of the identified findings along with their assigned level of importance.

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6.I DESIGN PROGRAM VERIFICATION The following sectionc describe the design program verification l J and resulting findings and corrective actions. This design i verification was performed by the Readiness Heview design verification team. The five members of the team had a cumulative experience of 85 years in power plant design engineering. Approximately 750 manhours were expended performing the verification.

)

i The design verification concentrates on the design interface l (i.e., the flow of design information) between Westinghouse, the l nuclear steam supply system (NSSS) vendor, and Bechtel Power

, Corporation (BPC). The results of the review provide a basis on

()

which to determine whether interface activities have been properly controlled.

Section 6.1.1 summarizes the verification results.

Section 6.1.2 describes the design verification scope and plan.

Section 6.1.3 describes the verification and results.

Section 6.1.4 discusses the design verification findings and responses. Section 6.1.5 discusses the significance of the

- findings.

4 6.t.1 SUMMAltY The scope of this module involved 11 key NSSS areas. The key I f areas and the verification sample reviewed ate presented in Table 6 .1 - 1.

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The design verification resulted in four findings. Three of the

! findings were procedural in nature and did not adversely affect hardware. The fourth finding identified hardware deficiencies.

The affected hardware will be reinstalled to correct

requirements. The finding which impacted hardware was i considered 1.evel 1. The remaining threa findings had no effect i on the technical adequacy, design, or hardware of the NSSS; nor do they indicate a deficiency in the overall NSSS interface between Hechtel and Westinghouse.

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The Level 1 finding identified a deficiency in a single discipline in which vendor mounting details for instruments wete not incorporated into instrument installation drawings. The project-proposed corrective actions to the finding, if properly implemented, should ensure that the deficiency will not recur.

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6.1.2 VERIFICATION SCOPE AND Pl.AN The Module 16 design program verification concentrated on the design interface between BPC/GPC and Westinghouse, the NSSS supplier. As noted in section 4.1, the Vogtle NSSS is designed by Westinghouse under the Westinghouse quality assurance J

b program. Westinghouse design work is organized using a functional group concept which is an organization divided into h; specific work activities, and in which a given type of work iu  !

done by a single group for all projects. Therefore, the same '

policies, procedures, 7-d often the same people are involved in a given design area for all plants. The Westinghouse quality assurance plan has been accepted by the Nuclear Regulatory Commission (NRC) for generic application to all safety-related work conducted in conjunction with the design of commercial h nuclear power plant projects. Repeated reviews by the NRC Quality Assur.nce Branch and the Region IV Vendor Compliance Branch have developed extensive documentation of the adequacy of the Westinghouse quality assurance program. In addition, numerous audits by Westinghouse customers over many years, including the 1982 Southern Company Services (SCS) Vogtle audit- h on design control, have established the acceptability of the Westinghouse quality assurance program. Because of this evidence of adequacy of the internal design process at Westinghouse, the verification scope concentrated on a revtow of the NSSS design interface.

Key areas were selected for review to ascertain whet.her the NSGS interface has been appropriately implemented and controlled.

These areas are:

o Piping stress analyses; o Reactor coolant loop equipment supports; o Westinghouse proof of- design review:

o Accident analyces; o Instrument and control interface o NSSS and bal -e of plant (bop) process requirementu, o Electrical interface; o NSGS equipment qualification and nozzle loads; o Radiation shielding source data:

o Equipment installation requirements; o Field change documents.

To maintain continuity between modules, the samples for verification were primarily taken from the residual heat removal l (RHR), auxiliary feedwater (AFW), and nuclear service cooling I water (NSCW) systems which were also the prima ry sa.aple systemn 1 in Module 4. Additiona~ NSSS systems such as the safety ,

injection system (SIS) and components such as reactor coolant g 6 . 1. - 2 f

loop equipment were also included to ensure that representative camples were selected for each of the areah in addition to the review of the design interfaces, l implementation of Final Safety Analysis Report (FSAR) I commitments was reviewed for those commitments associated with the NSSS which are unique to Vogtle (i.e., not associated with the Westinghouse NSSS generic design). The comnitments reviewed O are related to Westinghouse scope of supply and/or design.

Additionally, samples of commitments which represent Westinghouse design requirements on BPC design were reviewed to ascertain the implementation of interface of commitments.

(These commitments will also be verified in appropriate BPC design modules, such as Module 4 for equipment and piping requirements.)

The verification program also includes a review of previous audit findings, industry problem reports, and past project problem reports. The design interface was also reviewed specifically for the RHR/ component cooling water (CCW)/NSCW i system to verify Westinghouse /BPC design consistency.

l 6.1.2.1 Review Method The design verification was performed in two phases. In phase 1, licensing committaents which are unique to VEGP (i.e.,

not Westinghouse NSSS generic design) were reviewed to ascertain

() their implementation in project design documents. In Phase 11, design documents were reviewed to ascertain whether required l

design interface data has been properly transmitted, received, and implemented.

6.1.2.L.1 Phase I implementation of licensing commitments was evaluated through review of the licensing commitment documents (e.g., FSAR) and various implementing documents (e.g., design criteria). The commitments reviewed were those commitments related to the Westinghouse scope of supply which are considered unique to VEGP. Those commitments which are typical for Westinghouse O plants simi1.ar to VEGP are considered generic and were not specifically reviewed for implementation, since the Westinghouse NSGS design is performed oy functional groups using the same procedures and practices for each NSSS. The generic commitments are an integral part of the Westinghouse NSSS design.

In addition to the above commitments, a sample of commitments which represent Westinghouse process requirements on HPC- designed systems was also reviewed for implementation in BPC design documents to ascertain whether commitments / requirements have been appropriately communicated and implemented.

O 6.1-3 a

6.1.2.1.2 Phase 11 In Phase II of the verification, the design interface between Westinghouse and BPC/GPC was reviewed. Eleven key areas were selected for review to ancompass the major NSSS interface activities. These areas include each Bechtel discipline which is involved in interface activities and the areas address the flow of information in both directions through the interface (i.e., Bechtel to Westinghouse and Westinghouse to Bechtel). f Samples of design interface items were selected for review in each of the key areas. These samples included data provided to Westinghouse by BPC and data provided to BPC by Westinghouse.

The samples were primarily selected from the sample systems (AFW, ItHR , and NSCW). The review was directed as much as possible toward changes in design which would result in interface activities. To accomplish this, the review also included samples of systems (e.g., SIS) and components (e.g.,

reactor coolant loop equipment) for which design changes have occurred that required additional design interface activities.

Additional bases for the samples selected are discussed in the section 6.1.3.

Past audit findings, problem reports, etc. were also reviewed.

For this review, several examples were selected to ascertain whether project coordination and action had taken place. In particular, samples of Westinghouse-related 10 CPR 50.55(e) reports were reviewed for proper completion.

6.1.3 VERIFICATION HEVIEW The Westinghouse /BPC/GPC interface activities were examined for proper and effective exchange of information between organizations. The areas considered were transmission of information required by other parties, receipt and correct internal distribution 01 tr? information, implementation of information, and feedback, when needed.

The interface review was accomplished using a plan developed to systematically guide team members through their review.

Adjustments were made to the plan when the verification team found that the selected items were inappropriate or when more significant items were identified. The original worksheetc for the plan are shown in Figures 6.1-1 and 6.1-2. The worksheets were modified and appropriate documents identified for review after discussions with personnel from the organization initiating the information.

The following sections discuss the review and results of the selected interiace areas listed in section 6.1.2.

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l 6 .1. t .1 Pipe Stress Analyses ,

iO The liechtel/ Westinghouse interface in the pipe stress analysis area was reviewed for proper control and effective exchange of data required by each organization and for feedback of information. The review of the interface in this area concentrated on design changes which required interface activity. The following areas were selected for review to encompass most of the interface activities in this area:

o Large and small bore ASME 111 piping fatigue analysis; 4 o Seismic spectra; i

o Support loads / location; I o Jet impingement loads.

Design changes in the sample systems (RHR, AFW, etc.) were selected for review and additional samples were included to ensure a review in each of the above areas.

For piping fatigue analysis interface activities, changes to the RHR/515 injection lines to Heactor Coolant (RC) Loops 1 and 4 I

to the RC Loop lines 1201-036-12 and 1201-119-3, to SIS line 1204 02-12, and RC Loop 1 resistance temperature detector (RTD) manifold piping were selected for review. The current routing, including changes shown on thr7e isometric drawings and the

} current stress analysis input for the RHR/ SIS injection lines, was observed to be appropriately included in the Westinghouse fatigue analyses.

Reviewed were routing changes which were made following the completion of the current Westinghouse fatigue analyses to large bore ASME 111 Class 1 lines 1201-036- 12, 1201-119-3, and 1204-021-12 shown on 12 field change requests (FCRs) and their associated design change notices (DCNs). The appropriate i Westinghouse fatigue analyses were reviewed and found not to '

include these changes. However, it was determined that Westinghouse will review and finalize fatigue analyses prior to turnover. As a rule, the fatigue analyses are updated during the interim only if the routing changes are significant. In the O case of the above 12 FCRs, the routing changes were minor and did not require interim Westinghouse fatigue analyses.

Also reviewed was the Westinghouse fatigue analysis for small bore ASME 111, Class 1 RC Loop 1 RTD manifold piping. The current Bechtel input data and the addition of a tee in the

() piping were appropriately incorporated in the fatigue analysis.

The Westinghouse analysis produced unacceptable results. The transmittal informing Bechtel of this and of possible solutions was reviewed. The reviewer determined that Bechtel had redesigned the line, using a Westingnouse recommended solution and is now performing the stress analysis of the line.

6.1- 5

The seismic spectra being used by Westinghouse was reviewed.  !

This seismic spectra was observed to be the same as the current Bechtel seismic spectra given in Hechtel design criteria.

Changes to loads on four supports and a change of support location were identified i> discussions with a Westinghouse engineer. A letter informing Bechtel that these changes occurred was identified by Westinghouse. The letter could not be located in the Bechtel files. Iloweve r , it was determined that periodic reviews were made of correspondence files, and I these reviews. coupled with design interface meetings would have ensured that this letter would be located.

Jet impingement loads on the pressurizer surge line were selected for review. The Bechtel transmittal of the complete set of jet impingement loads (from all high energy line breaks) and another transmittal of pipe insulation data were reviewed.

The transmittal of the results of the Westinghouse jet impingement evaluation was reviewed. The pressurizer surge line was appropriately addressed in each transmittal.

The interface in the pipe stress analysis area was observed to be properly controlled and data effectively exchanged. There were no findings in this area.

1 6.1.3.2 Primary Equipment Support. Loads The review of the interface in the area of primary equipment support loads on the containment structure concentrated on changes in the loads for the steam generator supports and reactor coolant pump supports. In September 1982 Westinghouse revised the support loads from those"of December 1976, which were initially transmitted to Bechtel for the steam generator supports and the reactor coolant pump supports. The Bechtel calculation for the s*eem generator support loads and the calculation for the react"r enolant pump suppett loads were reviewed. Both calculations clearly summarized the evaluation of the respective equipment support loads up to and including the September 1982 loads. In April 1985, Westinghouse again revised the support loads for this equipment. These new loads g were recently transmitted to Bechtel. The reviewer, while in W the Bechtel offices, was shown that the new loads were being prepared to be routed in the civil discipline for evaluation.

The interface in the area of primary equipment support loada on the containment structure was observed to be documented and i

controlled. There were no findings in this area.

6.1.3.3 Westinghouse Proof-of-Design Review l

Verification of the Westinghouse proof of-design interface review was to be initially a review of the drawings and 6 . ~1 6

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i revisions used by Westinghouse in perf or ming the proof- of- design review against the present drawing and revision level for the

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1 safety injection and residual heat removal systems. Bechtc1 and )

Westinghouse engineers identified and the reviewer confirmed i that the Westinghouse letter transmitting results of the proof of- design review requested that ac- built revisions of the drawings use! in the review be provided to Westinghouse. This request had been assigned an action item number and was being

() tracked on the project open items list. Therefore, a mechanism was found in place to ensure Westinghouce review of changes to drawingc used in the proof- of-design review; thus, the originally planned review was not pursued. Since an as-built review is being tracked to completion, the interface activities in this area were found acceptable. There were no findings in this area.

6.1.3.4 Acci_ dent Analysis The review of interface in accident analysis covered three specific areas: containment subcompartment analyses, containment prensure/ temperature analyses, and emergency core cooling system (ECCS) analyses including revision to Bechtel input data and auxiliary feedwater flow requirements.

l In accident analysis, the primary area in which Bechtel utilizes data provided by Westinghouse is containment subcompartment analyses. A single calculation, steam generator compartment I pressure / temperature, was celected for ra"iew as representative j of the limited number of containment sur r.npa r tments . The calculation was observed to ae based on current Westinghouse mass and energy release data.

Westinghouse performs several types of accident analyses which require information from Bechtel. The containment pressure / temperature and ECCS calculations were selected for review since changes to the heat sink data, a portion of the Bechtol-provided information, had occurred.

The ECCS calculation was also selected for review since not only was information required from Bechtel to perform the t calculations, the calculation also defined the auxiliary feedwater system flow at various conditions which were required to be implemented by Bechtel.

The current Bechtel data- package transmittal for use in the containment pressure / temperature calculation was reviewed. The latest Westinghouse containment pressure / temperature calculation O was observed to be based on the current Bechtel data-package. A Dechtel transmittal of revised containment heat sink data, whi,u supercedes a portion of the current data-package, was reviT i The Westinghouse response to the revised heat sink transmit. t was reviewed. Westinghouse evaluation of the revised heat sink O

6.1 7 l

1 data determined that there was an overall increase of heat sink and Westinghouse reanalysis wat not required.

l The current Bechtel data-package transmittal for use in the ECCS calculation was reviewed. The latest Westinghouse ECCS calculation was observed to Le based on the current Bechtel da ta- package. Correspondence regarding iterations of the layout of the Bechtel-provided AFW system were reviewed. The layout iterations were to obtain the required AFW system flow at various conditions. The final of the six Bechtel and Westinghouse letters reviewed verified that tor the appropriate conditions, the AFW uystem flow would be that required by the Westinghouse calculation.

Both Bechtel and Westinghouse have controlled and documented the use of input data provided by the other for use in the analynes for which each is responsible. There were no findings in this area.

6.1.3.5 Instrumentation and Control Interface The interface in the instrument and control area was reviewed for incorporation of functional requirements specified by Westinghouse for Bechtel designed systems, and for the input signals to be provided by Bechtel to the Westinghouse solid state protection system (SSPS), which is a part of the engineered safeguards features actuation system. The selected sample is summarized i n Ta b l e 6.1- 1.

The Westinghouse functional requirements were reviewed for those which would be implemented by Bechtel in the instrumentation and l controls area, those which had changed or were added during the life of the project, and those which were a7sociated with the sample systems. Three c.pecific functional requirements for Bechtel-designed syst"me were selected for review. The AFW turbine pump start on b1ec?mut and steam generator blowdown and sample line isolation requirements of these were selected since they were added during a Westinghouse revision to a functional requirements document. Seven Bechtel control logic diagrams were reviewed. The logic for each required function was observed to be shown appropriately on control diagrams.

The reviewer was able to identify just two instrumentation and controls input signals required to be supplied by Bechtel to the Westinghouse SSPS. The main steam (MS) stop valve position and turbine trip system pressure signals were selected for review.

The Bechtel Solid State Protection Cabling Block Diagram was reviewed for these input signals. The four required channels of the MS stop valve position input signal and the three required channels of the turbine trip system pressure input signal were observed to be shown appropriately on the block diagram. The input signals were appropriately supplied by Bechtel.

6.1 - 8

j The review in the instrumentation and controls area showed that the requirements specified by Westinghouse were incorporated

() appropriately into the Bechtel design. There were no findings in this area.

( . l . 1. 6 NSSS and BOP Process Requirements The interface in the area of Westinghouse process requirements

was reviewed in two areas. The area of Westinghouse-designed fluid systems was reviewed for changes to the flow diagrams 3

initiated by Westinghouse to ensure that the changes were

reflected accurately in Bechtel drawings. The area of process 7s requirements for BPC-designed systems was reviewed for implementation of Westinghouse process requirements from a

(' Hatance of Plant Functional Requirements document and the Steam 1 Systems Design Manual. Table 6.1-1 eummarizes the selected j j sample. j Geven changes to the Westinghouse flow diagrams for four uafety related systems were selected for review. The Westinghouse letters transmitting these changes were reviewed. 1 Seven Bechtel process and instrumentation diagrams (P&lDs) were J reviewed for implementation of the selected flow diagram I changes. The P&lDs in each instance accurately incorporated the flow diagram changes. As evidence of an active, effective, and controlled interface in this area, two additional activities were observed. First, documentation was reviewed of a January O,

s 1984 meeting between Bechtel and Westinghouse engineers to discuss and resolve Bechtel comments on Westinghouse flow diagrams. Second, Bechtel and Westinghouse engineers working in 4

this area were informally meeting at the Bechtel offices at the time of the review. They discussed the work in this area and provided clarification for the reviewer on specific questions.

Gamples of Westinghouse process requirements for liechtel. designed systems which interface with the NSSS were selected based on a review of the Westinghouse Balance of Plant Functional Requirements documents and the Steam Systems Design  !

Manual. Eight requirements for the sample NSCW, CCW, and AFW '

systems were selected for review. Except for NSCW heat J

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exchanger redundancy, two separate flow paths, and AFW single failure criterion, the design requirements were selected for review because they had potential for being overlooked during the design process.

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%ix process requirements f or Bechtel-designed NSCW and CCW

( systems were selected for review from a Westinghouse Balance of Plant Functional Requirements document. Requirements for NSCW heat exchanger redundancy and two separate flow paths were observed to be accurately implemented in the Bechtel NSCW System Design Criteria. The limit on CCW heat exchanger maximum outlet temperature was documented by a Bechtel calculation. NSCW and

(}

CCW heat loads from the NSSS were reviewed in two separate  !

6.1 - 9 f , _ - _ _ , , . , y , , , , , ~ ..-,r- -

Bechtel calculations. These heat loads were the same as the current heat loads from Westinghouse vendor documents. The requirement for automatic start of the standby NSCW train was observed to be properly implemented in a Bechtel control logic diagram. The pressure requirament for the CCW piping to the reactor coolant pump therral barrier was observed to be correctly implemented in the Line Designation List.

The two process requirements for the Bechtel- designed auxiliary O AFW system were selected from the Westinghouse Steam Systems Design Manual The requirement for steam from two steam generators to the AFW turbine pump was observed to be properly implemented on a P&lD and on piping isometric drawings. The AFW system single failure criterion requirement was accurately implemented in the Bechtel AFW System Design Criteria.

The review of Westinghouse process requirements showed that the process requirements specified by Westinghouse were incorporated into the Bechtel design and that changes made by Westinghouse to the flow diagrams f or Westinghouse- designed systems have been controlled through the interface and are being implemented by Bechtet. There were no findings in this area.

6.1.3.7 Electrical Interface The interface in the electrical area was reviewed in three areas: engineered safeguards features actuation system (ESFAG) valve and pump train asrignments, Westinghouse electrical requirements f or Bechtel- designed systems, and input signals to be provided by Bechtel to the Westinghouse SSPS which is a pa r t.

of the ESFAS. The selected sample is summarized in Table t.1 1 Six pairs of valves, safety injection pump suction and discharye isolation, residual heat removal pump suction and discharge isolation, boron injo tion tank outlet isolation and sump suction 1.ine isolation, cr? cafety injection and residual heat removal pumps were reviewed for consistent train assignments between Bechtel and Westinghouse. This area was selected for review in part because of the team's awareness that problems had been experienced in the area on other projects. Westinghouse train assignments were identified in the Westinghouse Electrical Requirements document. Sixteen Bechtel elementary diagrams, one for each component, were reviewed for the Bechtel train assignments. The train assignments reviewed were consistent between Bechtel and Westinghouse design documents.

The Westinghouse Steam Systems Design Maneal was reviewed to identify electrica~ power requirements for the Bechtel designed systems. Two electric power requirements for the sample AFW system were selected for review: AFW motor-driven pumps powered from separate power sources and the AFW turbine- dr iven pump fl ow path independence from diesel AC power. The Bechtel element a r y wiring diagram for each AFW motor-driven pump and for each of 6.1-10

three pairs of valves in the associated flow paths were reviewed. The motor-driven AFW pump and associated valves in

.) each flow path were observed to be powered by a separate power source (train) from that of the pump and valves of the other flow path. The requirement for separate power sources for each AFW motor-driven pump flow path therefore is met.

1 f- The AFW turbine- driven pump flow path independence from dienel l j AC power was reviewed by identifying AFW turbine-driven pump l flow path valves, pump turbine steam supply valve power sources, j the source of AFW turbine- driven pump lube oil pump power, and independence of the lube oil cooling source. The power source f or f our AFW turbine- driven pump discharge line valves and one

,f 3 suction line valve is shown on the elementary wiring diagram for  !

'q ) each valve as an independent DC power source (train). The power source for two valves in the steam supply line to the pump turbine is shown on the elementary wiring diagram for each valve as an independent DC power source (train). The power supply for two vendor supplied, steam supply line valves was traced through the vendor drawing to a Bechtel one-line diagram which showed it as an independent DC power source. The AFW pump vendor manual showed that the lube oil pump is integrally attached to the AFW pump shaft and is not dependent on diesel AC power. The AFW pump lobe oil cooling system is internally cooled as shown on the vendor drawing and reflected on two AFW P& ids, and is also independent of diesel AC power. Thus, the turbine-driven AFW pump system was observed to meet the requirement of independence U

/~} from diesel AC power.

The engineered safeguards features (ESP) pump motor loads shown ,

on the Bechtel diesel generator loading table drawing were l reviewed against the Wes tinghouse- specified motor horsepcuers and were consistent.

The reviewer identified two electrical input signals required to be supplied by Bechtel to the Westinghouse SSPS. The reactor coolant pump under frequency and reactor trip breaker open signals were selected for review. The reactor trip breaker open signal was initially' required by the Westinghouse functional diagrams and subsequently was deleted by a revision representing a change in design. Four Bechtel elementary diagrams were

,['s reviewed. The reactor coolant pump under frequency signal was d--) shown as a separate channel on each of the four drawings. This signal had been implemented appropriately. The same drawings were reviewed for inclusion of the reactor trip breaker open l l signal channels which had been deleted by Westinghouse. The signal was not on any of the reviewed elementary wiring diagrams f which was consistent with the Westinghouse drawing. The input

\- uignals were supplied by Bechtel.

The train assignments of equipment and valves and the ESF pump motor horsepowers used in the Bechtel diesel generator loading table were consistent between Bechtel and Westinghouse design

}

documents. The Westinghouse- specified electrical power 6.1-11

requirements on Bechtel-designed systems were observed to be implemented. The input signals provided by Bechtel to the SSPS reviewed were observed to be implemented. The interface in the l electrical area was found to be effective and adequately controlled. There were no f;ndings in this area.

NSSS Equipment Qualification 6.1.3.8 f

Nuclear steam supply system equipment qualification was reviewed in three areaF; seismic qualification of equlpment, environmental qualification of equi pment , and nozzle loadings on equipment from the attached piping. The samples selected for review were electrical and mechanical equipment (including various types of each) and valves. For both the seismic and environmental qualification areas, the electrical equipment sample included the SSPS cabinets and four instrumentation transmitters from three different manufacturers. The valve samples were a modulating valve in the centrifugal charging pump portion of the chemical and volume control system (CVCS) with its associated electronic control module, and the safety injection pump discharge isolation valve with its associated motor operator and limit switch. For mechanical equipment, safety injection pumps were reviewed in the environmental qualification and nozzle loadings areas. The HilR heat exchanger and steam generator main steam nozzle loadings were selected for review. The sample was selected to review the seismic qualification, environmental qualification, and nozzle loadingu for the same equipment reviewed in other areas to the same extent possible. The samples for each area are summarized in Table 6.1-1.

6.1.3.8.1 Seinmic Qualification The review in the sei mic qualification area was intended to verify the seismic qual.fi ,t i on for safety-r,, elated NSSS equipment to Vogtle-specific seismic requirements.

Bechtel was observed to ha e on file the current revision ot ,

Westinghouse documentation related to seismic qualification for l each piece of equipment of the sample selected for review in the  !

seismic qualification area.

1 The review of seismic qualification of the selected sample at electrical equipment could not be performed since the program for comparison of seismic qualification levels to Vogtle-specific seismic levels for electrical NSSS equipment wan I still under development at the time of the review. Based on this, it was observed that there was a lack of specific definition of such a program in the Bechtel Project Hefetence  !

Manual. Finding 16-15 was issued as a result of this observation. 1 6.l. 12

1 The two sample valves were reviewed for seismic qualification.

The Hechtel stress calculation was reviewed for the line in

() which the modulating CVCS valve is located. This stresc calculation defines the Vogtle-specific seismic acceleration levels for the modulating valves and references the Bechtel Stress Analysis Criteria-Design Criteria document on a standard format sheet. Although the Design Criteria states a valve acceleration limit as a pipe support design goal, there was no (s) direct comparison of the resulting specific calculated acceleration to the modulating valve qualification. The electronic control module for this valve is mounted separately and would be treated as electrical equipment.

g-) The safety injection pump isolation valve seismic qualification acceleration levels were reviewed in the Westinghouse equipment q_j specification for the valve. The acceleration levels of the installed valve were reviewed in the stress calculation for the line in which the valve was located. The calculation was

' observed to contain a preprinted sheet referring to compliance

...with the criteria established by Westinghouse," but there was no direct comparison of the qualification to calculated Vogtle-specific seismic acceleration' levels. The reviewera observed that the Bechtel-calculated acceleration levels were within the qualification levels of the Westinghouse equipment specification. Finding 16-12 was issued dealing with the lack of direct comparison of qualification to calculated Vogtle-specific acceleration levels for the two valves reviewed.

The stage of work progress in the seismic qualification area did not permit a sufficient review to draw conclusions regarding this area. Review of the valve seismic area indicated weaknesses in documentation which did not affect the technical adequacy of the work.

6.1.3.8.2 Environmental Qualification l The review in the environmental qualification area was to verify

( the environmental qualification for saf ety- related NSSS l equipment to Vogtle-specific environmental conditions.

Bechtel had on file the Westinghouse documentation, in its current revision, for each piece of equipment of the sample selected for review in the environmental qualification area.

The comparison of environmental qualification conditions with location specific environment was in progress at the time of the I~' review. A systematic program was being followed in which a System Component Evaluation Work (SCEW) sheet was prepared for NSSS safeti related electrical components and electrical parts of mechanical components and valves which operate in a harsh environment.

O 6.1 13

i Completed SCEW sheets were reviewed for the modulating CVCS l valve and associated control module, the safety injection pump discharge isolation valve motor, the four instrument ll transmitters, and the safety injection pump motor. These GCEW sheets properly comparo' the 'omponent environmental qualification conditions .t'. the location-specific environment for each item, indicating the component was qualified. No SCKW sheet was observed for the safety injection pump discharge isolation valve limit switch. Bechtel project personnel showed lh the reviewer that the limit cwitch was schedulad on a list of work to be cowpleted.

A systematic program was also being followed for safety related mechanical equipment located in a harsh environment. The Mechanical Equipment Qualification (MEQ) List identifies the l l qualification of soft part components (gasket, seal, etc.) that could degrade. The safety injection pump and the safety injection pump discharge valve itself were observed on the MEQ list. The documentation of the soft parts lifetime qualification for the valve was observed to be satisfactory during the review. The safety injection pump will be evaluated, because it is included in the MEQ, when Bechtel receives Ibe qualification data.

The Bechtel equipment and instrumentation qualification list (EIOL) makes a comparison of saf ety- rela ted equipment, environmental qualification conditions and the component l

location specific environment for both harsh and nonharsh environments. The environmental comparison for the SSPS cabinet was observed in the EIOL to be satisfactory.

Completed SCEW sheets were located for all sample equipment except one valve limit cwitch which was shown by project personnel to be on the list of work to do. The SSPS cabinets are not located in a harsh environment and do not require a SCKW sheet. The environmn't for the SSPS, as defined in Design Criteria Documents 100F, -"i onmental, is compared to the Westinghouse qualification in a Bechtel internal tracking program, ElQL. The safety injection pump had been identified by the project for later verification of soft parts (any parts of l the pump whose function can degrade as a result of environmental exposure qualification).

Although work was still in progress in this area, a complete program exists to p r ov icie a comparison of environmental qualification conditionr with loca tion- specific environment a l requirements Co. appliccble equipment. There were no findings in t. h i s area.

6.1.3.8.3 Nozzle Loadings Equipment nozzle loadincs from attached piping were reviewed for three different types of Wostinghouse supplied components.

6.1-14 l

l l

Selected for review were the main steam nozzle for Steam Generator' Number 1, the inlet nozzle for RHR heat exchanger A, and the suction and discharge nozzles for both safety injection pumps. For the steam generator and RHR heat exchanger, the allowable nozzle loading specified by Westinghouse in each component's equipment specification was met by the nozzle loading from the attached piping calculated in the Bechtel stress calculation for lines 1301- 001- 26 and 1205 005-8

.()

respectively. The Bechtel-calculated loadings on the four safety injection pump nozzles exceeded those of the Westinghouse safety injection pump equipment specification. The Bechtel letter requesting approval of the '3fety injection pump calculated nozzle loads and Westinghouse's subsequent letter of 7

approval were reviewed. l

'k The interface activities in the area of nozzle loadings are l controlled and data is effectively exchanged. There were no findings in this area.

6.1.3.9 Radiation Shieldinu Source Data i l

The review in the area of radiation shielding source data  !

concentrated on the use of current Westinghouse data in the l Bechtel calculations. Three Bechtel radiation shielding calculations were selected for review: volume control tank (VCT) valve gallery, Auxiliary Building Level D (RHR pump shield f-wg wall), and Auxiliacy Building Level D post-accident shielding

( ,/ calculations. The current and past revisions of the

, Westinghouse Radiation Analysis Manual were reviewed to identify I areas where data had been revised. The VCT valve gallery and Auxiliary Building Level D (RHR pump shield wall) calculations were selected for review since Westinghouse data which had been revised should be used as input to the calculations. The Auxiliary Building Level D postaccident shielding calculation was selected to review a calculation for other than normal l operating mode shielding. l The Bechtel VCT valve gallery calculation was observed to use i appropriately the data from the November 1978 issue of the '

Westinghouse Radiation Analysis Manual.

5 i

The purpose of the Bechtel Auxiliary Building Level D (RHR pump shield wall) calculation review was to adjust the previous calculation to account for changes that occurred in the November 1978 issue of the Westinghouse Radiation Analysis Manual since the previous issue. The Westinghouse data was observed to be appropriately utilized in this calculation.

}

The Bechtel Auxiliary Building Level D postaccident shielding calculation was observed to be based on the Bechtel source spectrum for a TMI shielding calculation which was observed to be based, in turn, on the Bechtel source terms for the TM1

(~ Ghielding Calculation. During a review of the source terms for V) 6.1-15

1 the TML Shielding Calculation, the appropriate use of data from the November 19'/6 Westinghouse Hadiation Analysis Manual was obqerved. h During the review of tb Westinghouse Radiation Analysis Manual.

it was observed that the cettent manual was issued in March 1983. The changes made in the March 1983 issue were observed to be typographical and had been previously identified h to Westinghouse by Bechtel. Bechtel calculations were not affected by the typographical changes in the March 1983 inuue of the Westingho",e Radiation Analysis Manual in Bechtel calculatiens and, therefore, no changes to Bechtel calculations were required.

The review of Bechtel shielding calculations showed that the O current Westinghouse radiation source data had been appropriately used. There were no findings in this area.

6.l.3.10 Equip _ ment Installation Requirements The interface in the area of equipment installation requirement u was reviewed in the electrical area and the control systemu area. Based on the team's experience, installation of mechanical equipment was judged an area which typically receiven much attention on a project and usually has more easily identifiable requirements. The team considered the installation of electrical or control equipment to be an area in which installation requirements were more likely to be misinterpreted.

The installation requirements for nuclear instrument system (NIS) cable were selected for review. Westinghouse provides the installation requirements in the NIS Cable and Connectors Installation Contrn! anc Electrical Systems (C&MS) Standard document. The Bechtc. taceway Systems Construction Specificetion references t. r% E S Standard document and reiterates specific details from the C&MS Standard document The Bechtel taceway drawings which contain NIS cable were observed to call out specifically the Raceway Systems Construction Specification for NIS cable installation details process control instrumentation installation was also selected for review. The sample for review was selected to include various types of Class 1E transmitters (e g. pressure, level) from different manufacturers (e.g. Barton, Tobar) and to include installations inside anc outside of containment. The containment pressure, accumulator level, steam generator wide range level, refueling water storage tank (RWST) level, pressurizer presstire a n c? steam generator pressure transmittern were selected for review.

A Westinghouse drawing showing the containment pressure transmitter, a sealed transmitter system, was reviewed for 6.1 16

installation details. The Becntel installation drawing for this transmitter was reviewed and ooserved to reflect accurately the

() installation details of the West inghouse drawing.

The Uechtel instrument installation drawings for the accumulator l level and pressurizer pressure transmitters were not prepared at i the time of this review and could not be reviewed.

() The three Westinghouse transmitter drawings, one each for the steam generator wide range level. RWST level, and steam generator pressure transmitters were reviewed to identify installation details. The Bechtel instrument installation drawing for the steam generator pressure transmitter was also reviewed. The torque requirements for the transmitter bracket

() mounting bolts, shown on the Westinghouse drawing, were not included on the Bechtel drawing. The Bechtel instrument installation drawing showing installation details for both steam generator level- wide range and RWST level transmitters were reviewed. The transmitter bracket mounting bolts shown on the Hechtel drawing were 1/4 in. bolts, not the 5/16 in. bolts shown on the Westinghouse drawing. Based on the discrepancies between Westinghouse transmitter installation details and the two Bechtel instrumentation installations drawings, Finding 16-13 was issued.

In the areau reviewed, except for the transmitter mounting which l tesulted in a finding, the Westinghouse installation 1

("% requirements were reflected in Bechtel design documents. I

( l 6.t.1.lt Field Chanqe Documents l The field change documents reviewed were deviation reports '

(DRu), field change requests (FCRs), and field change notices iFCNs)/ field equipment change orders (FECOs) 6.1 3.Li,1 Devialion Reports in the review of 1000 DRs in Module 4, only nine were identified as being for Westinghouse equipment. Four of those nine were

/'

reviewed as part of Module 4 and were revealed to have been i processed with proper Westinghouse interface. Because of the l smal1 number of DRs found in Module 4 that were related to Westinghouse equipment, only three more were reviewed for this module. The three DRs reviewed were on the No. 1 steam generatot supports. the tout safety injection accumulator tanks, and the No. 1 tegenerative heat exchanger. All three had been O- reviewed by Westinghouse, and that review was documented on the DR. The interface activities for DRs were found to be controlled properly.

O 6.1- 17

1 l

l 6.1.3.11.2 Field Change Requests 1 Field change requests reviewed as a part of the Module 4 verification indicated that FCRs were being processed by Bechtel in accordance with appr med rocedures. Based on this, FCRs were not explicitly reviewad in this module. FCRs were, however, a part of the review of the pipe stress analysis interface area as discussed in section 6.1.3.1.

lh 6.1.3.11.3 Field Change Notices / Field Equipment Change Orders Westinghouse- issued FCNs were reviewed to determine whether they were being issued to construction and being incorporated into llechtel FECOs for construction implementation. The FCNs reviewed were selected from four 10 CPR 50.55(e) reported issues in order to determine whether cootdination of related j commitments had taken place. The four issues cover Westinghouse 3 inch gate valve closure, Westinghouse 4 inch gate valve closure, Limitorque pinion key shear, and Limitorque valve opetator qualification.

For the four issues, Westinghouse identified four corresponding FCNs issued for Unit 1. The four FCNs were processed properly, having been received and issued through BPC Drawing and Document Control. Four FECOn for the respective four FCNs were reviewed dnd proper issue for construction was verified.

Although they had been issued for construction, it was not apparent from the FCNs or FECOs that the respective modifications were a rerult of a material nonconformance/

deficiency. The reviewer also could not identify a specific tracking method to ensure the implementation of commitments related to 10 CFR 50.55(e) issues. This resulted in Finding 16- 11 6.1.4 DESIGN PROGRAM VERIFICATION FINDINGS During the performance of the verification activities described 1 in sections 6.1.2 and 6.1.3, questions were raised which l required clarification and resolution, or there were findings of deviations from commitments or procedures which required project evaluation, disposition, and corrective action. Each question or issue identified during verification was documented using the l Readiness Review finding procedure. A Finding Report form was l completed and processed. During the verification process for '

this module, four findings were identified. Table 6.1 - 2 provides a summary of those findings.

A detailed description of these tindings and the bauis for conclusions regarding them are as follows:

l 6.1 18 N

o Finding 16-11 (Level II)

~I ) Finding: In reviewing commitments related to ks/ 10 CFR 50.55(e) reports to the NRC for file M-30 Sheared Pinon Keys in Limitorque Motor Operators, and file M-47, Westinghouse Gate Valve Position Indication, the following conditions were identified:

)

[ 1. Documented evidence could not be located to verify th&t either 10 CFR 50.55(e) commitment had been completed.

2. It could not be determined whether these reportable deficiencies had been documented on Deviation i() Reports, or other nonconformance documentation.

3. It does not appear that there is a program in place to track completion of commitments to the NBC regarding 10 CFR 50.55(e) reports.

Project Response: The following items address the

, respective items in the finding.

1. A review was conducted of two 10 CFR 50.55(e) reports identified in the finding'wich the following results

M30 - Sheared Pinion Keys in Limitorque Motor Operators It was determined that field equipment change orders 1

(FECOs) were issued to replace the pinion keys in 12 Westinghouse-supplied valves identified in the report. However, there ware no change documents issued for the 50 BOP valves identified in the report.

I j

Omission of the BOP valves was an inadvertent oversight. The 10 CPR 50.55(e) report was issued  ;

because of a material deficiency identified in l Limitorque models SB-0-25 and SMB- 4 operators. It I l

if'T was determined that these two models were not used on Vogtle. However, while evaluating the pinion key iV material concern for reportability. SCS licensing noted that Limitorque motor operator models SB-0, S B- 00, SMB-0, and SMB.00 all used the same pinion key material as the SB-0-25 operators which had failed.

Tests had shown that this material was adequate for  !

(' use in these operators, but audits conducted by

\ Westinghouse and Bechtel found weaknesses in Limitorque's material control programs such that a weaker resulfurized carbon eteel material could have been inadvertently installed in any of these operators, as had happened in the cases

O 6.1-19

of failure involving the SB-0-25 operator models.

It was determined that it would be more effective to consider the concern reportable pursuant to lh 10 CFR 50.55(e) and replace the pinion keys in all BOP SB-0, S".00, "MB-0, and SMB-00 Limitorque operators than to conduct material analyses on all the pinion keys. However, the decision to replace the pinion keys in the other Limitorque operator was not relayed from licensing to engineering, and lh engineering did not recognize the commitment to rcplace the pinion keys when they received the 10 CFR 50.55(e) report.

M47 Westinghouse Gate Valve Position Indication h

Wiring modifications to provide positive indication of valve closure were committed to the NRC 10 CFR 50.55(e) report X7 BG0 3- M4 7, GN-280, dated November 9, 1983. Required corrective actions to meet this commitment were implemented and tracked through CCP B10159M In addition to the two items above, reviews were also conducted of several other evaluations involving commitments to the NRC:

M15 and M17 3-in. and 4-in. Gate Valve Closure Problem M44 Brown Boveri Heactor Coolant pump Switchgear M72 GE AKR- 30 and AKR- 50 Circuit Breakers 1&E Bulletin 83 Nonconforming Materials Supplied by Tube-Line Corporation In all cases L t.a corrective actions were being

( tracked by appropriate change documentation.

! Therefore, the discrepancy regarding the sheared pinion keys is considered an isolated case.

2. The program for 10 CFR 21 and 10 CFR 50.55(e) l reports does not require that associated material l

deficienciet be documented on Deviation Reports.

Operations Deficiency Reports, or other nonconformance documentation. These documents are used by conrtruction and operations personnel respectively to identify nonconformances discovered in the field or in documentation. Reportable deficiencies identified at the engineering / design level would not normally be documented on DRs/ODRs.

Bechtel s Deficiency Evaluation Report (DER) is used to identify and track these deficiencies.

Reportable 6eficiencies resulting from 10 CFR 21 6.1-20

l l

l reports from vendors and subvendors are also )

identified on DERs or other engineering or licensing i I\ documents while they are being evaluated. The

\- resulting corrective actions are documented on FECOs ,

or Change Control Packages (CCPs) until completed. I FECOs and CCPs are tracked to ensure they are completed and signed off when required.

l

[) Wording in Revision 12 of construction procedure

\' G D- T-- 01 inadvertently indicated a broader scope for DR usage than was intended. A revision is currently being processed to more specifically define the use of DRs within the Vogtle Project's nonconformance control program. This revision will be issued by l

() November 15, 1985.  !

3. The project had previously recognized the need to i track corrective action to items reported to the NRC I and, as a result, had established the Project Compliance Coordinator position. The Project Compliance Coordinator is responsible for immediate  !

followup and assignment of individuals for ,

accomplishing corrective actions. Since the establishment of the Project Compliance Coordinator.

position, the primary emphaais has been to track the most recent items reported to the NRC. Readiness Re /in W Pinding 16-11 has pointed out the need for j gg additional tracking, especially for the older 1

( j/ reportable items. By November 26, 1985, the following actions will be completed:

i

a. Review all Vogtle Project 10 CFR 50.55(e) l reports Cot corrective action commitments.

b. Obtain objective evidence of completed l corrective actions.

c. Update existing tracking logs as required.

In addition to the above, GPC QA has developed a computer tracking system for all commitments made to the NRC relative to 10 CFR 50.55(e)/10 CFR 21

((~N_) reports and responses to NRC inspection reports and 1&E Bulletins. This system will be used by QA in the long- term f ollowup of corrective action commitments and will also be available to the Project Compliance Coordinator and others as a cross- checking device.

, Readiness Review

Conclusion:

Inplementation of corrective action to review all past reports to the NRC, coupled with the commitment tracking and followup, should ensure that no additional commitments are

~T overlooked. The FECO and CCP do provide a means of (G

6.1-21

tracking resulting design changes through construction completion.

h o Finding 16-12 (Level 111) i Finding: Inadequate reference in the dynamic streus calculations to allowable valve acceleration levels prevents a valid comparison of calculated versus allowable accelerations for Westinghouse-provided f valves. Two sample valves in two calculations were revieved to confirm that appropriate Westinghouse allowables were used. Although Design Criteria Document 1017 is referenced by a preprinted form contained in the calculation packages, DC- 1017, revision 3, dated December 6, 1984, does not explicitly contain the h current allowables for one of the two sample valves.

Appropriate specifications or qualification reporte are not referenced. Both calculations conclude that the effective accelerations calculated are acceptable without reference to an appropriate acceptance criteria source.

. Pro.iec_t H e s g o n s e.:

1. 1K3- 1208 485 The allowable valve accelerations for llV-190A given by specification X6AA06-512 are not currently shown in DC- 1017 (revision 3).

DC-1017 was revised to incorporate specification X6AA06 512.

2. 1K3- 1204 020- 03 : Until June 1982, Bechtel utrens calculationc used a format that did not give a numerical comparison (of valve accelerations). In June 1982, r stress group instruction memo changed the format of the valve acceleration qualification sheet. '

T5is format provides the numerical comparison tar 1ource reference. The calculation for 1K3- 1200 028 03 was originally done in 1980 and revised in 1981. This calculation was subsequently transferred to the jobsite in 1983 and the engineering group supervisor (EGS) signature added in 1984. Since a technical revision was not done after 1981, the new format for valve acceleration qualification was not included. A technical revision will be done by V- SAMU f o r the as- built reconciliation and will include an actual / allowable comparison.

g Extent: Bechtel llome Of fice Engineering ( BIIOE ) I conducted a rev'.ew and ensured all other valves requi r i ng noism:_c qualif ication are listed i n DC- 1017 1

l 6.1-22 I

1

- t

! Impact on Hardware:

• l t

t

) There was no adverse effect on hardware becat ;e calculated g values are below allowable values.

l Hoot Cause of Finding:

l g~g 1. DC-1017 did not incorporate allowable valve accelerations for all saf ety- related NSSS and BHOM

(} valves. This was an oversite by BHOE and is considered an isolated case, based on the above review.

2. Stress calculations performed prior to June 1982

() used an old format which did not give numerical comparisons or source reference.

l Action Taken to Prevent Recurrence:

1. Specification No. X6AA06 512 was incorporated in revision 4 of DC- 1017 signed off August 9, 1985. A review of the FSAR by BHOE concluded that no other allowable valve accelerations were missing from DC- 1017.

l 2. Stress group instruction memo No. 11 was issued in

! June 1982, requiring updated format for all new

' fg calculations and calculations to be revised after l (_/ June 1982. During routine as-built reconciliation, l acceleration comparisons (actual / allowable) will be

{ included in calculations that do not already include this comparison.

Headiness Review

Conclusion:

Tae new format for Bechtel stress calculations will ensure comparison of valve seismic accelerations for seismic qualification when the as- bui l t reconciliation is completed. Based on the review conducted by BHOE, accolarations for all valves requiring seismic qualification are included in DC- 1017. The response is acceptable.

o Finding 16-13 (Level I)

The installation bolting details, bolt size, and torque values for Westinghouse-supplied seismically and environmentally qualified electronic dp transmitters and pressure transmitters are not the same on the Bechtel drawings as those on the Westinghouse drawings. For the

! -) selected sample of three specific transmitters reviewed.

l two Bechtel seismic mounting detail drawings apply.

Installation bolting details on neither Bechtel drawing were in agreement with those of the corresponding Westinghouse drawing.

6.1-23

l 1

l.

l l

Proiect Response: Instruments 1FT-922 and PT-501 are ,

shown on installation isometrics 1X5DY00922- A and 1X5DY00501-A respectively. The Instrument Installation Drawing referenced in these isometrics is CX5DPM030. We concur that the nrobloms listed in the finding are applicable to installation drawing CX5DPM030.

Extent: A review was conducted of the CX5DPMXXX drawings which are provided to mount Westinghouse and lh l other vendor's safety-related instruments (project class l 61J or 11J). This review was performed comparing the mounting requirements given in the vendor documents and the requirements as shown in Instrument Installation Drawing. Of the 21 drawings which are used to mount 61J l and 11J instruments, 7 were found to include all vendor l mounting requirements. The remaining 14 drawings showed l one of the inconsistencies as noted in the three l

categories below:

A. Vendor recommended torque value was not shown on

! the drawing.

B. Incorrect bolt size called out on the drawing.

C. Combination of both A and B.

Impact on Ilardware: According to information provided by Pullman Power Products, as of July 1, 1985 there are 1

66 class 61J and 11J instruments installed and tested of I which 64 are affected by the results of this finding.

Using the categories of inconsistencies given above, the breakdown consists of A=47, B=8, and C=9.

Root Cause of Finding: In the initial issue of the Instrument Installation Drawings, certain vendor l requirements 'or instrument installation were inadvertently omit:'d from review and therefore were not incorporated. This omission has been determined to be i due to an oversight by the designer and checker.

1 This finding is unique to Control Systems' mounting ,

drawings since these drawings are designed as stand alone documents, not requiring the vendor print for installation. Other disciplines rely on the vendor l print, and do not duplicate the vendor's information. 1 l

Action Taken to Prevent Recurrence: Instrument l

Installation Drawings will reference the vendor I

documents which contain vendor requirements for equipment mounting. The checker will check these references against the completed Instrument Installation Drawing to assure that all requirements pertinent to installation are incorporated into the drawing. These I l 6.1-24 l

i actions will be contained in a Desk Instruction to be f- issued by November 22, 1985.

U In addition, training will be conducted with designers and checkers to ensure that all installation requirements pertinent to vendor's equipment a qualification are recognized and the source documents I are identified.  ;

Future Commitments: The 14 Instrument Installation Drawings identified have been revised to incorporate the vendor's qualification requirements. In addition,  !

reference to applicable vendor prints and instruction manuals for 61J and 11J instruments has been added to j() the 21 Instrument Installation Drawings used for I installing these instruments. The affected instruments will be reinstalled and tested in accordance with the requirements of the revised installation drawings.

Readiness Review

Conclusion:

The reinstallation of affected instrumentation and revision of affected drawings will resolve the problem identified. The design verification team concurs that the problem was isolated in the Control Systems discipline. The response is acceptable, o Finding 16-15 (Level II)

Finding: VEGp Project Reference Manual, part C, Sectton 37, Equipment Qualification, does not adequately outline the verification procets for seismic qualification of NSSS equipment. The text does not address the activity entitled, Compare Requirements with Results, as shown in the block diagram, Attachment II.

Seismic formats are not included, and a verification procedute is not defined. The role of Westinghouse, EpC, or others is not discussed. Based on the above lack of information, it could not be established that a l complete program exists for verification of NSSS equipment seismic qualification.

Proiect Response: Westinghouse is responsible for

( O. generic qualification of NSSS equipment and provides to Bechtel the installation details consistent with their

, generic qualification program. This information is used l

by Bechtel in the design of anchor bolts and supports to ensure that the mounting design with which the equipment

('] is qualified is consistent with mounting details on

\/ installation drawings. Deviations from Westinghouse requirements during design and installation are reviewed by UpC Westinghouse concurrence is obtained as necessary. Bechtel and Westinghouse jointly prepare the NRC form Seismic and Dynamic Qualification Summary and Status of Safety Related Equipment for all NSSS 6.1- 25 l

L .

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i saf ety- related electrical and mechanical equipment (including pumps and valves). This list identifies necessary references which contain information regarding qualified and applicable seismic levels to enable BPC to make comparisons between qualified and actual location seismic levels. This form is submitted to NRC for review prior to their seismic audit. Code A in the qualification status column indicates that the equipment has been properly qualified for VEGP requirements.

The Be htel Equipment Qualification stQ) Group (with support from the Westinghouse Seismic Group) is responsible for ensuring that NSSS equipment is properly

, qualified and meets VEGP requirements. Completion of l the above effort is documented in the VEGP EQ Data Packages (EODFs). These EQDPs are reviewed, checked, and signed off by the VEGP task force which consists of the Bechtel EQ Group, SCS licensing, GPC nuclear operations, and the Westinghouse seismic consultant.

These EODPs are controlled project documents and will be turned over to Georgia Powr:r Company for use during plant operation.

1 toot Cause of Finding: Personnel involved in the abovo l process are familiar with the requirement and did not l feel that a detailed description of the above l

description was necessary, i

Future Commitments: PRM Section C37 will be revised to include the above details and will be issued by December 2, 1985.

Readiness Review

Conclusion:

Inclusion of the equipment seismic qualification program details in the PRM will satisfactorily resolve the finding. The response is acceptable.

6.1.5 FINDINGS SIGNIFirANCE l

The four design verification findings are summarized in Table 6.1-2. Each finding is classified into a level of importance to plant safety as defined in the introduction of section 6.0.

1 One of the findings was Level 1. The Level 1 finding related to lack of inclusion of vendor mounting details on Bechtel drawings used by constrection to install qualified transmitters. The g

l T l Bechtel project response to this finding, when completed, should resolve concerns addresced by this finding.

Only one category, seiunic, was established where two of the findings bore a relationship. The definition in the Bechtel PitM &

of the details of the Bochtel equipment seismic qualification W 6.1-26

4

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program should resolve any cumulative concern for this t ca t eirory.  ;

,9 Collectively the findings do not affect the adequacy of the NSSS l interface between Bechtel and Westinghouse or associated design i

! activities.

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l TABLE 6.1- 1 (SHEET 1 OF 4)  !

REVIEW SAMPLE

SUMMARY

( )  :

( })i pe S t r e s s A na ly_s i s o Reactor coolant loop RTD manifold bypass line O o HHR recirculation / safety injection, hot leg injection line o Pressurizer surge line O Primary Equipment Support Loads o Steam generator supports o Reactor coolant pump supports l

Accident Analysis o steam generator compartment pressure / temperature o Containment pressure / temperature 10

(,) o Auxiliary feedwater flow requirement ,

o Emergency core cooling Instrumentation and Control o Function requirements

1) lli-Ili steam generator level close feedwater i

isolation valves i 2) Blackout signal start turbine driven auxiliary feedwater (AFW) pump

3) Actuation signal for AFW motor driven start also O closes blowdown and sample lines for all steam generators o Solid state protection system input signals

1) Main steam stop valve position O- 2) Turbine emergency hydraulic fluid pressure 1

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l 004 5 p/ 323- 5/2 l

TABLE 6.1-1 (SHEET 2 OF 4) b M NSGS _a nd BO)) . Process Hequirements o Flow diagram changes f()

i 1) 2)

Reactor coolant system Safety injection system l'

3) Containment spray system

4) Chemical and volume control system

! o Process requirements O

l 1) Redundant nuclear service cooling water (NSCW) heat l exchangers i 2) Two separate NSCW flow paths

3) Automatic start of component cooling water (CCW) j backup pump on running pump stop l

4) CCW piping to reactor coolant pump thermal barrier designed for 2485 psi

5) Maximum CCW heat exchanger discharge temperatute limited to 1200 F

6) NSCW and CCW heat loads

7) Auxiliary feedwater (AFW) turbine driven pump supplied steam from two steam generators ir"}

(b

8) AFW system designed for single active failure

,E l e c t r i_c a l o Trai n assignments

1) Valves 8801 A & B

2) Valves 8802 A & B

3) Valves 8809 A & B

4) Valves 8811 A & B

5) Valves 8812 A & B l 6) Valves 8923 A&B l 7) Safety injection pumps 1 & 2

! 8) Residual heat removal pumps 1 & 2 o Power requirements

1) Motor driven auxiliary feedwater (AFW) pumps power from separate power sources which meet separation

[~ requirements

2) Turbine driven AFW independent of both motor driven

! APW pump power sources

/~%

0045p/323-5/3 1

l TABLE 6.1- 1 (SHEET 1 OF 4) i o Solid state protection system input signals

1) Heactor coolant pump under frequency

2) Heactor trip breaker open (deletion)

I) l NSSS_Xquipment Oualification I l

o Seismic and Environmental Qualification

1) Modulating valve located in the centrifugal pump

) portion of the CVCS; I l( Electronic control module, mounted separately l 2) Safety injection pump discharge isolation valve l Limitorque operator j - Limit switch

3) Solid state protection system cabinet l 4) Flow transmitter (mfg. by Tobar) located outside containment

5) Flow transmitter (mfg. by Veritrak) located inside j-% containment

l.  ;

j 6) Pressure transmitter (Barton) located outside i

containment

7) Pressure transmitter (Veritrak) located inside containment o Nozzle loading

1) Hesidual heat removal heat exchanger

2) Safety injection pump

• i

3) Steam generator (main steam)

Ghie ldirtg Source Da ta o Volume control tank valve gallery o Auxiliary building level 0 - HHH pumps o Source terms for TM1 shielding

• Envitonmental qualification also.

0045p/323-5/4 f

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TABLE 6.1-1 (SHEET 4 OF 4) .

r-i installation Requ_irements .

o Nuclear instrumentation system cable o Instrumentation - transmitters i(f-l' 1) Containment pressure (sealed)  :

2) Accumulator level l

3) Steam generator level l 4) Refueling water storage tank level l 5) Pressurizer pressure l( 6) Steam generator pressure ,

i Deviation Reports I

o 03528 No. 1 Steam Generator s u p po r t.s o 07067 Safety injection accumulator tanks (4) o 06759 No. 1 Regenerative Heat Exchanger l

l l

Vi. eld Change.. Notices (FCNs)/ Field Equipment Chance Orders (FECOs).

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10 CFR 50.55(e) Issue FCN Number FECO Number MIS & M17 Gate valve closure GAEM 10527 N-0022- 0BF ML5 & M17 Cate valve closure GAEM-10549 N-0045-0BF M10 1.imitorque pinion keys GAEM-10562 N- 0058- 0BF M69 bimitorque operator GAEM 10574 N-0071-0BF l qualification I

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004bp/323 5/5

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TABLE 6.t-7 RLADI NESS REv i E's F I NDI NG SUMfMRY finding Resolution / Conclusion /

Number Finding Level Project Response Assessnent 16-! ! Lack of documentation that convii t- 11 Tracking program being Response ments related to 10 CFR 50.$$(e) instituted. Past and Acceptable reports have been completed or are present reports to NRC to being tracked. be reviewed and corrective .

actions tracked 16-l? 1.ack of documentation for 111 Documentation will be Response corrperisons made of val we acceleration included per existing Acceptable levels. (1982) calculation format during "as-buiit" reconciliation.

16-13 Inaccurate qualified process i Drawings are revised and Response transmitter seismic mounting bolt affected transmitters will Acceptable details. reinstalled. Training of designers will be done to incorporate vendor ,

qualification details on Instrument installation Drawings.

16-15 1.ack of defined complete program to 11 Existing program to be Response verify NSSS equipment seismic included as part of the PPJ4 Acceptable qualification.

0067p/2

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• 1.0 NftC Commitments for NSSS VEGP-FSAR For cheqklist items 1.1 and 1.2' Select a mintansa of ten (10) carmitments identified on attactument A and verify that the corunitment is adequate and complete in the ispiementing docuruent.

1.1 Procedures, instructions, specifications AllSI f145.2.

11 Section 2 1.2 Design Criteria ANSI N45.2.

11 Section 3 i

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i Figure 6.1-1 Design Readiness Review Checklist (Sheet 1 of 11) i

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Bechtel/ Westinghouse (R-W) (M-B) l Georgia PWR/ Westinghouse (GPC-Il) (W-GPC) a) Identification of interface b) Responsiblittes i

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i Figure 6.1-1 Design Readiness Review Checklist (S hee t 4 of 11)

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Figure 6.1-1 Design Readiness Review Checklist (Sheet 11 of 11)

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B-W HW ixx.u,en t west ingtruse Doctsnent Number farne nt s item No. Inter f ace Itum g TItie Nurter il ti e B-w i n t t er RTD Manifold A-l Ganges to HQ. RTU Stress Analysi s Bypa ss Mani f o ld Piping B-W totta Pressurizer Surge BW-Jet Impingnent Loads U-w talculation A-2 Line Stress on Pressurizer Surge-Ana lys t s line B-W La t ter (Lexp i) Stress Calc

/ RtH Recirc. and SIS Refs.

tbt Leg injection A-3/4 M Stress Calc Stress Analysis input B-W te tt er (Low 4)

Refs, L

BW A-5 Large Bore Routing B-W Oange Line Line Line Selmic Spec tra OW B-W Sei sm i c Spec tra A-6 Transni tia l of Selsulc Book Spectra Stres s Calc, Cha nge s B-W Revi si on A-7 Suppcri Load Chsngos V I -120 4-Supper t 025-6001 W-0 Sw por t St ress Ca lc Support 025-tOO2 W-B Supgort 025-tD12 W-B Swport 047-H003 W-B A-8 Sepor t Locat Ion Gange W-B i50 Yl-1204-051-H028 W-B Ca lc.-Je t Let tw A-9 Pipe Break Locat lons RtfySi Loop 3 _ Impin gsent 0049V/2 Figure 6.1- 2 Design Verification Work Sheet (Sheet 1 of 16)

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t< - n bit Ocx. cont mo st ing hw se Doceent Numter iltie Nurrter Comnent s item No. Inter f ace item y 11 t i o W-B Structural Stepor t Ana lysi s B-l Strean Generator (S/G)

Support Loals Ca Icu I at Ion SeI smic Ancher Mot ton W-8 is-6 Struc ture t Smrort Analysis B -2 Change to S/G Ca lcul ation Rev Stoport Loads St r uc tur a l Sup por t Ana ly s t s B-3 Rertor Coolant Pump Stoport loads W-B Ca lcul at ion Sel smi c Anchor Hat lon W-e Structural Suppor t Ana 1 ys t s B -4 RP Stpport Loads Ca Icu I at ion Rev Cha nge s 4

0049V/3 Figure 6.1-2 Design Verification Work Sheet (Sheet 2 of 16) ,

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uRT AINHL N1 PHLSTRE/TLHTRATURL ANAL YS 15 AmIDEHi 44ALYS6S B-w OPC Docunent West inghouse Document W-B Tltie Number Tltie Numter Conment s iten No. Inter f aco it em Mass and Energy Releams W-0 Ca lcu l at lon Letter

1. D-l W-0 Calculation letter for Contairvnent Sub-compartnent Analysis B-W Ca Icu I at Ion D-2 ContaIrvient P/T and ECCS Backpressure Calculation Calcu Iat lon input data Ca lcu l at lon Ca lcu l at i on BW-D-3 New test sink data in B-W contalrvnent disposition?

D-4 AfW f low requ irenents W-B D-W i

0049V/8 Figure 6.1-2 Design Verification Work Sheet (Sheet 4 of 16)

O O I&C INTERFEE B-w BFC Document West Inghouse Document item No. Int s face item W-0 Tltie Hunt >er iltie Nueer Conenent s E-1 High-high steam W-0 Logic Diagram functional generator weter ieve1 RequIremonts in two of four (2/4) level channels in any of the four steam generat<rs (P-14) should trip the main turbine anf all neln feedwater pums, close l the nein and bypass j feedwater control valves for eli steam generators and close all feedwater Isolat ion valves in series with the main and b'/ pass fee < heater corttrol va lves (see Section 8.17.1.3 for f urther ca lar i f i cat lon) .

E-2 A black-out signal W-0 Logic Diagram Func t ional shnold start the Rewirements turbine driven auxi l l er y f eud pug .

E-3 Any autoneted notar W-0 Logic D1agram Functlonal driven auxiilery feed Requirements pump actuat Ion signal i should close the blow- f I

doun Isolet lon ard sample line valves for all steam generators.

1 0049V/9 Figure 6.1-2 Design Verification Work Sheet (Sheet 5 of 16)

O O O stC INTEFF E E O O O O B-W BPC Docunent West Ingtu;use Docunent Nur@er Conument s item No. Inter f ace item M Tl ti e Number Title E -4 Input tn SSPS Mein stean sicp valve W43 Drawing logic ulagram p3sition or Logic Diagree Turbine enerpncy trip W -0 Ora ul ng Logic Diagram systta nydrquilc fluid or Logic Diagran presstre signals

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0049V/lO I

Figure 6.1-2 Design Verification Work Sheet (Sheet 6 of 16) l

o al ST INr.HaJSE f NOCI.S 5 F1QUIHEMt.Ni5 NSCW B-W OPC Ducument West inghouse Docunent item No. I nt er f ace it en w-B Tltie Number il ti e Number Commen t s F-l Since the (N5CW) tw at W-ti tesign Cri teria DC- t unc t iona a exchangers are pwt of Requi reent s engineered safegurds, suf ficient redundency Is required to meet single f ailure cr i ter ion.

F-2 liedur:1ency requirements W-0 Design Cri ter i e DC- Functional placed on the (NSCW) Requi rment s system design by the single failure criterion, require that two separate flow paths be provided for all engineered safeguards systems components.

F-3 If delng normal W-B Design Critwla DC Functional operation of the plant Requirenents the (MSCW) CCS punp becomos in< perat i ve, Instrunentat lon erst I controls automatically start the backup l

pump.

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0049Vfi l Figure 6.1-2 Design Verification work Sheet (Sheet 7 of 16)

WESTINGtWSE PRO &SS REQUIREMENTS NSCW West Inghouse Docunent B-W BPC Document Tl ti e Number Comments Numeer lien No. Inter f ace item y Tltie Design Criteria DC Functional (MSCW) cogonent cooling W-B i F-4 Rept rements system piping must be designed to withstand pressures 12485 psigl resulting af ter Isolation with (RCP) thermal barrier In leak age. >

Design Critarla DC Functional The seximum permissible W-0 F-5 Re@lranents or (MSCW) CCS heat endenger dIseharge Ca Ic. _

temperature during cooldown is specifled as i200F.

W-0 Design Orlterla DC Functional F-6 NSOf heat loads ard Regirements Tables flows I

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i" 0049V/12 g Figure 6.1-2 Design Verification Work Sheet (Sheet 8 of 16) i

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,NGHOISE PROR55 HEQUIREMENTS G

wf FEW B -W DPC Document West Inghouse Docunent Huseer Comments Itam No. Interface Item g TItie Number TItie F-7 For stese turbine W-0 Design Criteria DC Stem Systems driver (MW ptsap), Design Menual stese must alemys be avellable from at least two stese generators during plant geration, to preclude a loss of all steau sgplies due to any single inc Idont .

F-8 The (MW) system must W-B Design Crlieria DC Steam Systens be designed to Design Manual sustain a single act i ve f ai l ure in the stort term liess than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) without loss of ' flow capabillty belou the ministan re

tw Cherated Valves a. Design &lteria DC to provide water to S/G

b. M>ttr (berated Valves b. Design &lterla DC ,

in steam sipply

c. Lube oil punps c. Design Criteria DC l d. Lube oil cooling d. Design Criteria DC sipplied by an or external systesa. Procurement Spec.

(Use of the turbine t o cm l l uba oi l driven punp nut:Is t h i :. ) . ' G-Il Rectcr (bolant Pump W-8 Drawing Logic Diagram ! Steply underfregtency or  ; j signal to SSPS Logic Diagram IG-12 Reactor trip breaker W-0 Draw ing Logic Diagram cpen signal to SSPS- or deletion Logic Diagram i

• I l oo49V/16 Figure 6.1-2 Design Verification Work Sheet (Sheet 12 of 16)

SEISMIC QtALIFICATl04 B-W BFC Document Westinghouse Document item No. Interf ace item Q 1111e Nun 6er Titte Number Comment s H-l Valve HCV-190-A u -t3 Seismic Qualsitcallon Revlev laam Form Electronic (SQRT) Control pt>dule HY-190-A W-0 SQRT H-2 Valve 8802A M) tor W4 SQRT l (perate Limit Sul tch for 8802A W-B SQRT Valve 8802A W-B SQRT H-3 Velve B-W BW-Exceed "g* or end loads B-W DW-H-4 SSPS Osblnet W-0 SQRT H-5 FT9t B (Tobar 1 (OC) W-0 SQRT FT 444 (Veritrak) (IC) W-B SQRT PT 418 (Barton) (OC) W-B SQRT PI4SS (Ve'itreAI r (IC) W-O SWI H-6 R>ft heat occhenger Nozzie loads B-W BW- Ca lcu i at lon H-7 S1 Pung Norzle Ioah B-w BW- Ce seu Iat Ion , H-8 Stesa Generator Main Steen pbzzle B-W DW- Calcu lat ion Loeds I Figure 6.1-2 Design Verification Work Sheet (Sheet 13 of 16) 4 9 O O O O O O E NVIIU4ME N T 4. Ol#18 F I CAT l 04 i i D-w DFC Docunent West Inghouse Doctment l 4 ltere No. Interf ace item g Title Number Title Number Comment s k i l-l Velve HCVl90 W4 Sub-Component l Electronic Control W-B Evalust lons Worksheet (S&W) ' Module HY-190 SaW l 1-2 Valve 8802 A mtor W-B S&W Operator , Limit Switch for 8802% W-B S&W valve 8802 A W-B SGW 1 -3 SSPS W-B S&W 1-4 FT 918 W-B S&W FT 444 W-B SGW i PP 418 W4 S&W PT 455 W-B S&W I-5 Rtet Host Exchanger W-B S&W Tag ' 1 1-6 St Pump W-B Saw Tag _ J I ' 'l 0049V/18 Figure 6.1-2 Design Verification Work Sheet (Sheet 14 of 16) , SHIELD 503RCE DAT A B-w BFC Docunent West inghouse Document item No. Inte f ace itorn Q Title Ntruber Title Number Conument s J-l volume (bntrol Tank Sotrees Rad iat ion Analysis W-6 Shield Sizing Radiat ion Analysis Design Manuel Rev. 2 Calculation Design Manual to Rev. 3 ( Attached) for Room RA40 or 80 J-2 Rm Punp Room wall W-0 Shleiding Ca lc. Radi ation Analyses at passageway Design Manual e I 0049V/19 Figure 6.1-2 Design Verification Work Sheet (Sheet 15 of 16) ~ _- . - ~ _- . , . . . . - . . . . . . _ _ . . - . . . -. .. . . - - - . ~ . . . - = _ . .. . . ~ , _ _ _ . . _ .-._ - - INSTA.LAT IO4 REQUIREMENTS B-W BFC Docisnent We st ingtouse Doctment Nun 6er Iters No. Inter f ace I tern Q TItie Number TItie Conument s K-1 NIS Cable and Cbnnectors W-6 installation Specification CAES Standard K-2 instrument at lon W-0 insta l lat ion Spec if icat ion Dra ulng installation I l OO49V/20 Figure 6.1-2 Design Verification Work Sheet (Sheet 16 of 16) l l 6.2 CONSTRUCTION PROGRAM VERIFICATION ] 1 Program assessment of the construction section of this module l was performed to provide an evaluation of the nuclear steam supply system (NSSS) installation and related construction j activities performed by Nuclear Installation Services Company 1 (NISCO). l 1 1 The Readiness Review Construction Team for Module 16 consisted l of five team members who have a combined total nuclear construction experience of 48 manyears. Program assessment represents approximately 675 total manhours expended by the . construction team. 6.2.1

SUMMARY

EVALUATION Program assessment activities resulted in five findings. The findings were reviewed by the construction team to assess their ,

impact on the project and to classify each finding with respect l to the following categories in addition to levels of importance:

l o Category A, Hardware: )

o Category B, Pap ^rwork: I o Category C. Pts grammatic.

The findings, their level of importance, and their category are given in Table 6.2-1.

The construction team did not discover any hardware-related findings during verification activities. The hardware surveyed demonstrated the acceptable installation of the NSSS. Four findings were classified as Category B, paperwork concerns.

These findings involved minor deficiencies in documents, procedures, or markings, and further review demonstrated that no hardware was affected. NISCO currently maintains approximately 1400 documents and procedures. The construction team reviewed a f- portion of approximately 800 of these during assessment activities, and only the 4 minor deficiencies were identified.

The construction team concluded that the documentation adequately reflects the acceptability of the hardware covered by Module 16.

The remaining finding was classified as a Category C,

( ) programmatic concern. The finding resulted from a deviation between the NISCO procedure and the Bechtel specification for the sequencing for the reactor internals assembly. The finding concerned deviations in the internals assembly sequence and the performance of leveling the upper internals assembly flange during intermediate preinstallation assembly activities.

() Westinghouse has verified that the sequence of the assembly activities in question is not critical. The upper internals assembly flange levelness has been checked and verified by NISCO

l t

as required by the specification. These intermediate activities had no hardware impact and are not significant to final internals assembly or installation.

l ~he construction assessment activities indicate that an O

acceptable program has been applied for fabrication and installation activit i ansseiated with the NSSS even though j findings were identified. The construction team concluded that the established program and N1SCO construction performance resulted in an acceptable system.

6.2.2 PROCLAM ASSESSMENT PLAN The assessment plan was developed to provide an appraisal of the NSSS installation performed by NISCO. Development of the plan l consisted of the selection of items and activities for l

evaluation, determination of the current status of these items and activities, and formulating the method (e.g., surveys, paperwork reviews, etc.) for assessing each item and activity.

l 6.2.2.1 Assessment Item Selection l

! The assessment areas selected are dividad into two parts, hardware / components and program / procedures. These areas provide a representative outline of the NSSS hardware and certain programmatic activities that support field installation. The basis for the selection was comprised of several factors.

First, only items or activities within the NISCO scope of responsibility were considered. Second, historical information such as audit reports, Nuclear Regulatory Commission (KRC) inspection reports, Institute of Nuclear Power Operations (INPO) evaluations, and NRC Construction Appraisal Team reports from VEGP and other similar nuclear facilities were reviewed to identify generic or significant problem areas that would warrant assessment. Finally, items or activities that have been, or will be, assessed in 'ther Readiness Review modules were identified to preclude the duplication of assessment activities. Examples of these areas are piping installation, covered in Module 4; pipe supports and pipe whip restraints, covered in Module ll; measuring and test equipment, covered in Appendix G: and instrumentation and controls, covered in Module 20.

Using the basis described above, the following items and activities were selected for assessment:

o Hardware / components f

- Reactor coolant pump No. 4 motor setting,

- Reactor coolant pump No. 4 support columns and tie rods.

- Steam generator No. 1, 1

1 1

6.2-2 l 1

- Pressurizer,

() -

Reactor pressure vessel (RPV),

RPV head assembly, l

- RPV internals,

- Bottom-mounted instrumentation; o Program /Procedtres

- Material control.

- Nondestructive examination, l

- Document / records control, ,

- Personnel certifications.

- Nonconformance handling.

O 6.2.2.2 Hardware / Components The selected hardware / components were assessed by a combination of witnessing in-process activities, walkdown and visual examination, and review of documentation. The assessment method details and the results of the hardware / component assessment are as follows:

6.2.2.2.1 Reactor Coolant Pump No. 4 Motor Setting The setting of the motor on reactor coolant pump No. 4 was Os assessed by the construction team to evaluate the following activities associated with the setting of the motor

l

! o Rigging; l

o Lifting:

l o Maintenance activities; o Documentation of activities.

The assessment ascertained that the activities were performed in compliance with specification and procedure requirements. The assessment also ascertained that required inspcetions, including maintenance inspections, were performed and appropriately documented. The checklist shown in Figure 6.2-1 details the specific attributes checked during assessment activities.

O' 6.2.2.2.2 Reactor Coolant Pump No. 4 Support Columns and Tie Rods The vertical supports and tie rod brackets of the reactor coolant pump No. 4 were assessed to ascertain whether they were installed in compliance with specification ASME Boiler and l(

l Pressure Vessel (B&PV) Code requirements.

l 6.2-3 l

The assessment items consisted of 3 vertical columns, column bases, and connections and the 3 tie rod brackets and locking l bars, and 34 of the 54 connecting fillet welds.

The assessment activities were performed using a checklist developed by the construction team (Figure 6.2-2). The checklist details thc ;pecitic attributes, items, and documents checked during assessment.

The assessment activities included a hardware survey of the supports to evaluate the following attributes: '

o Material identity; o Location and orientation; o Clearances; o Locking devices and spacers; o Visual weld examination.

The assessment also included a review of installation documents to assure adherence to NISCO procedures for installation document control processes. Examples of the attributes assessed are as follows:

i o Availability and legibility:

o Appropriate entr,ies and signatures; O o implementation of selected specification requirements.

The construction team concluded that the installation and inspections of the vertical columns and tie rods were in compliance with specification and ASME B&PV Code requirementu.

6.2.2.2.3 Steam Generator No. 1 The loop 1 steam generator and the four steam generator vertical supports were assessed to ascertain compliance with procedure, specification, and ASME B&PV Code requirements. The assessment was accomplished through paperwork reviews and hardware surveys to assure that installation activities, including required inspections are properly documented and adequate.

The hardware surveys assessed the following attributes:

o Steam generator orientation; o Material / component identification:

O 6.2 4

o Vertical support orientation and clearance; o Vettical support pin connections.

The paperwork reviews consisted primarily of a review of FISCO Process Control Sheets (PSCs). The reviews were performed to ascertain the following attributes:

o Availability and legibility; o Appropriate entries and signatures; o Implementation of selected specification requirements.

The assessment activities were performed using checklists developed by the construction team (Figures 6.2-2 and 6.2-3). The checklists detail the specific attributes, items, and documents checked during assessment.

The results of the assessment showed that the documentation was adequate to ascertain the acceptability of the installation of the steam generator. The hardware survey found the steam generator and vertical supports to be properly installed and identif.ied.

l l

'~^

6.2.2.2.4 Pressurizer The assessment of the pressurizer was to ascertain whether the l installation and setting was in compliance with procedure. '

specification, and ASME B&PV Code requirements. Additionally, the assessment evaluated the implementation of the quality control (QC) and quality assurance (QA) programs associated with the pressurizer.

The construction team assessment focused on four general areas:

o Field assessment; o Installation documentation; 9 o Procurement documentation; o General documentation.

The assessment was governed by the use of a checklist prepared l by the construction team (Figure 6.2-4).

The paperwork reviews revealed that the installation and setting of the pressurizer complies with construction specification X4 AZO 6. Observations of the N1SCO QA/QC program showed there was sufficient involvement to assure that the installation 9 requirements were satisfied. The field examination revealed that the basic orientation of the pressurizer is correct in

6. 2- 5

relationship to the reactor coolant loops and that it conforms to ASME B&PV Code requirements. The construction team identified one deviation concerning the stamping of the ASME nameplate. The deviation was documented on Readiness Review I'inding 16-8 and is described as follows:

o Readiness Ret.ew vinding 16-8 (Level II)

Description:

Pressurizer ASME nameplate does not indicate the Code class.

Proiect Response: Investigation of the finding shows that the NPT Code plate in question will be removed prior to N stamping. Westinghouse Quality Helease 24405, revision 1, shows that the stress report will be g completed prior to the primary hydrotest, and the W Westinghouse Authorized Nuclear Inspector (ANI) will authorize application of the permanent nameplate with the correct stamp and Code class after completion of the primary hydrotest and stress report.

Readiness Review

Conclusion:

Readiness Review concurs with the project response. The pressurizer Code class was verified as acceptable by review of the N-2 code data report.

6.2.2.2.5 Reactor Pressure Vessel The RPV was assessed to ascertain whether the setting of the HPV O and the installation of the RPV supports were in compliance with specification and procedure requirements. Since installation has already been completed, the assessment was done by reviewing the documentation associated with these activities. The checklist shown in Figure 6.2-5 developed for this review, outlines the specific attributes assessed by the construction team.

The construction team found all documentation to be administrative 1y and technically accurate and reflected the acceptable installation of the RPV and supports.

6.2.2.2.6 RPV llead Assembly O

The Unit 1 RPV head assembly was assessed through reviews of paperwork because the head assembly was not accessible for visual inspection. The paperwork reviews involved a review of the PCSs for the assembly of the RPV head, and were documented on Figure 6.2 5. The PCSs were reviewed for the following attributes:

o Legibi!ity and completeness of entries:

O 6 . 2 -- 6

g o Supporting documents, if applicable, attached or i referenced; w/

o Specification requirements properly reflected.

In addition, the documentation associated with seven head r~g penetrations was reviewed for additional attributes. The seven

(,) penetrations selected are listed on Table 6.2-2. The additional l

attributes assessed are:

o Material traceability; je s o Welder qualification; i

)

{'

o Inspector certification:

o Hydrotesting activities; o Performance of nondestructive examination (NDE).

The construction team visually examined 16 CRDM-to-RPV head penetration seal welds on the Unit 2 RPV head assembly to supplement the Unit 1 RPV head assembly paperwork reviews. The .

examination was performed using acceptance criteria established l by NISCO procedure ES-100-5, which are the same criteria used j for welding activities on the Unit 1 head assembly. The 16 i welds examined by the construction team are listed in b("N Table 6.2-2.

The results of the assessment found that the PCSs were complete and legible and demonstrated that the head assembly complies with specification and ASME B&PV Code requirements.

Additionally, the visual examination of the welding revealed the welds to be adequate and free from surface defects.

- 1 6.2.2.2.7 RPV Internals The RPV internals were assessed to accertain whether the l internals were assembled in accordance with specification

.[s requirements. The assessment was performed by review of installation documentation because the internals assembly was nearing completion and no assembly or installation activities were in progress. The documents reviewed were selected from the documents required to ascertain the implementation of 35 out of the approximately 70 specification requirements for assembly.

l[\w') These documents included NISCO procedures, installation drawings, and PCSs.

A checklist that details the specific attributes checked was prepared and used for the review of the documentation (Figure 6.2 6).

A D:

6.2-7

As a result of the assessment, the constructton team initiated Readiness Review Finding 16-6. This finding related to three instances where specification requirements were not adequately reflected on PCSs.

Readiness Review Finding 16 6, identified during assessment activities, is descri'.d b9 low:

o Readiness Review Finding 16-6 (Level II)

Description _: Of 50 technical specification requirements reviewed by the construction team. ' were found to not be completely implemented by NISCO procedures and/or PCSs.

Project Response: The two specification requirements in question involve the internals assembly sequence and the upper internals assembly flange levelness requirements while in the stand. These assembly activities have no hardware impact and do not affect final internals assembly and installation. This finding is not significant because the required assembly activities were completed even though the sequence was changed.

Additionally, the upper internals assembly flange levelness had not been completed at the time of Readiness Review assessment. This requirement has be'n e

checked and found to be within tolerance. It should be noted that this requirement is applicable when the internals are in the storage stand and does not affect the final installation. A Deviation Report was initiated to document these deviations and was dispositioned Use- As- 1s.

Readiness Review

Conclusion:

Readiness Review concurs with the project response.

6.2.2.2.8 Bot tom- Mv;c .ed Instrumentation The bottom-mounted instrumentation (BMI) was assessed to ascertain wuather the installation of the guide tubes and guide tube supports is in compliance with design requirements. In addition, the assessment ascertained the implementation of a portion of the N1SCO QA/QC program. lh The installation of the guide tubes and guide tube supports is currently in process. The construction team assessed completed portions of 6 of 58 guide tubes and 3 of 6 guide tube supports for the following attributes:

o Freedom from damage; o Installation per drawings; O

6.2-8

o Material acceptability and traceability; o Acceptable welding-visual examination.

To supplement the survey of the BMI hardware, the construction team reviewed the applicable installation documentation for the g- following attributes:

o Availability; o Legibility; o Supplemental records attached or referenced:

em)

V

o Appropriate entries.

The hardware surveys and paperwork reviews were performed using checklists that detail the specific attributes checked by the construction team (Figures 6.2-7 and 6.2-8).

The construction team found no deficiencies during assessment activities. Visual examination revealed that the welds checked were in compliance with approved criteria. All material checked was the required type, traceable, and all the documentation was retrievable, legible, and appropriata'y completed.

O

(, / 6 2.2.3 Procram/ Procedures 4

The assessment of program / procedures was performed in conjunction with hardware and component surveys. The assessment i

, consisted primarily of paperwork review activities but included l

visual examinations and interviews with construction personnel. '

The specific method of assessment used and the results for each area are as follows:

6.2.2.3.1 Material Control 2

Material control was assessed to ascertain whether the material H

' .) and components used by NISCO, including welding filler materials, were properly received and inspected, and that the documentation for the material was reviewed and accepted. The i

assessment also checked that welding materials were being .

controlled and issued in accordance with requirements. NISCO I has currently issued 162 Receiving Inspection Reports (RIRs) for I material and components and 26 RIRs for weld filler materials.

The construction team reviewed five of the material / components R1Rs and five weld filler material RIRs for the following attributes:

o identification of the material (type, size, tag no.,

.V

.O heat no.):

6.2-9

l o Appropriate signatures for inspections:

o CMTRs, if required, in conformance with specifications:

l o Evidence of acceptance or rejection.

The assessment activi..es were performed using a checklist that details the specific attributes checked by the construction team (Figure 6.2-9). The construction team found no deficiencies during the assessment. Documentation was readily retrievable, legible, and properly completed.

I I NISCO currently maintains two storage / issue areas for weld I fillet material and employs nine welders. The construction team

( surveyed one weld filler material storage area using the l checklist illustrated in Figure 6.2-10.

l l The construction team found the weld filler material to be i stored within the required housekeeping and environmental l requirements. Access controls were observed and the holding l oven was properly maintained and calibrated.

Three welders who had been issued welding filler materials were l also surveyed by the construction team using the checklist shown l on Figure 6.2-11. The assessment found that the material was issued in appropriate containers and was clearly identified, and a copy of the Welding Material Requisition was kept with the issued material.

O 6.2.2.3.2 Nondestructive Examination Nondestructive examination was assessed to ascertain whether NDE activities were properly implemented in accordance with project procedures. The assessment focused on the three NDE methods used by N1SCO: magnetic particle (MT), liquid penetrant (PT) and radiographic (RT). The assessment consisted of the review of completed inspection m acorts. Because of its frequent use, the assessment of the PT method was supplemented by witnessing the performance of PT examinations. Table 6.2-3 lists the specific items checked during assessment activities. The inspection reports were reviewed for attributes such as the following:

o Test part identification; O o Consumable materials; o Governing code:

o Technique data; o Examiner's qualification:

O 6.2-10

i

, 7- o Examination results:

o Legibility; o Completeness.

All assessment activities were performed using checklists

() prepared by the construction team (Figurea 6.2-12 through 6.2-15). The checklists detail the specific attributes checked.

Eleven MT inspection reports of the 18 completed were reviewed by the construction team. The assessment determined that the documentation was adequate to ascertain the acceptability of the

() inspections, but one procedural deficiency was identified on Readiness Review Finding 16-9.

Thirty-one PT examination reports of the 299 completed were reviewed during the assessment. The reports were determined to be adequate and complete. In addition to reviewing documents, the construction team witnessed the performance of two liquid penetrant examinations. The construction team found that both examinations were performed and documented in accordance with procedures and that the technique used was acceptable.

Radiographs were assessed by reviewing the radiographs and associated inspection reports for 35 ASME B&PV Code section III,

(~ subsection NF support welds. The review was performed by a GPC

(_}/ individual who was qualified and certified to the requirements of ASNT-TC-1A, Level III. The 35 welds reviewed included original and repair welds. As a result of the review, six comments were identified and addressed on Readiness Review Finding 16-10.

Readiness Review Finding 16-9 and 16-10, identified during assessment activities are described below:

o Readiness Review Finding 16-9

Description:

The MT reports reviewed during assessment do not reference the governing code for acceptance criteria.

Proiect Response: All of the 18 MT inspections performed were in accordance with ASME B&PV Code section III. NISCO has corrected the MT inspection reports to

, reflect the governing code.

\~) The PCS controls the work activities, including NDE, and indicates the governing code. The QA/QC engineer visually inspects the item and performs the NDE in accordance with the PCS: therefore, the inspectors were cognizant of the governing code and the examinations were conducted and interpreted in compliance with the k- appropriate acceptance criteria.

6.2-11

Readiness Review

Conclusion:

Readiness Review concurs with the project response.

o Readiness Review Finding 16-10

Description:

NISCO radiographic reports and film do not fully comply ..th ASME B&PV Code sections III and V.

Deficient conditions exist relating to:

1. The identification of specific revisions of NISCO procedures used for interpretation.

2. Incomplete sketches and inadequate identification of rejectable indications.

3. Lack of location markers or maps.

4. Differing dates on film and reports.

5. Not addressing linear indications.

Proiect Response: Items 1 through 4 of Finding 16-10 involved minor documentation discrepancies with the radiographic reports or the implementation of procedure requirements for completing radiographic reports. None of the discrepancies affected or questioned the quality of the weld. The radiographic reports were corrected as applicable. The radiographic examination report form and the radiographic shooting sketch form have been revised to include acceptance procedure revision numbers ll and locating dimensions of rejectable indications.

NISCO has also conducted additional training with NDE personnel on code and procedure requirements.

Item 5 of finding 16-10 involved five films of a similar joint configuration that displayed linear indications not noted on the radiographic reports. The linear indications  : e not addressed because they have been interpreted to be slag pushed up under the clip angle and out of the weld area, which is common with this joint configuration.

ASME B&PV Code film was re-reviewed to ensure that h linear indications have been addressed. This review has been completed, and the ASME B&PV Code film is in full compliance with Code requirements.

Readiness Review

Conclusion:

Readiness Review concurs with this response.

f 6.2-12

6.2.2.3.3 Document / Records Control The assessment of document / records control was performed in conjunction with paperwork reviews of NISCO PSCs. The assessment ascertained the following:

j (('T o Drawing numbers listed on the PCS were correct for the

,) time of installation / inspection; o Drawing revisions were documented on the PCSs:

o PCSs reflect appropriate information and entries; o PCSs were available and retrievable.

4 The construction team reviewed approximately 12 PCSs during assessment activities. All the requested PCSs were available and retrievable. The construction team did identify one deficiency:

o Readiness Review Finding 16-7 (Level Ill)

Description:

One PCS out of 29 reviewed eferenced drawing numbers but did not list their revision level as required by NISCO procedure ES-147.

,\/

[) Project Response: The lack of drawing revisions on the PCS had no effect on the quality of the hardware.

Procedure ES-146, Document Control, assures that only the latest revisions of client / engineer and

! NISCO-originated documents are available for use. In l addition, the PCS is prepared using current design and

installation criteria and is approved by NISCO QA/QC and j the client / engineer. These program controls eliminate the possibility of accepting incorrectly installed hardware because of a lack of documented procedure revisions.

The revision numbers of the drawings identified in the finding have been entered on the PCS. NISCO has

() reviewed an additional 25 PCSs and noted that the correct revisions had been entered. The personnel responsible for entering the drawing revisions on the PCSs have been reinstructed in procedure requirements.

Readiness Review

Conclusion:

Readiness Review concurs

()

~

with the project response.

6.2.2.3.4 Personnel Certification Personnel certifications of NISCO QC inspectors and NDE

[A technicians were assessed to ascertain whether inspection activities were performed by qualified personnel. The 6.2-13

l l

construction team reviewed individua'l certification packages for evidence that the personnel were certified to the proper level i using the method applicable at the time of performance of the activity.

The construction team reviewed five certification packages of N1SCO personnel. Thi aampia is representative of nine certified QC inspectors and NDE technicians employed by NISCO since the beginning of inspection activities. The construction team found that for all cases the inspection personnel were appropriately certified.

6.2.2.3.5 Nonconf ormance llandling An assessment of N1SCO nonconformances or DRs was performed to h I ascertain whether resolutions of DRs were in compliance with NISCO pr ocedur e ES- 14 2. The assessment involved 27 DRs that were reviewed by the construction team for the following attributes:

o Appropriate disposition approval signatures:

! o Proper closure and completion.

l The 27 DRs reviewed were representative of the 135 DRs that have been initiated by NISCO. The construction team found all DRs to i

be legible and appropriately dispositioned and completed.

6.2.3 COMMITMENT IMPLEMENTATION l

Section 3.4 contains the commitment matrix for the commitments identified by the FSAE that are applicable to Module 16. After identification of the commitments, Readiness Review reviewed each commitment to identify the project document that currently implements that commitment. Additionally, the review was performed to identit 3 the project document that initially implemented the commitment; .e., the project document in ettect when the work activity began.

The commitment implementation matrix identified 11 construction commitments. Each commitment was appropriately traced to implementing procedures from initial implementation to current status.

l 0029 p/325- 5 h

6.2-14 l

l

l t

l l

l Table 6.2-1 CONSTRUCTION FINDINGS O

Level of RRF Importance a [Jumber Category D ,_

_ _ , Description 11 16-6 C 1. Asnembly sequence for RPV j internals not in accordance with specification.

i 2. Flange leveling requirements i e- not implemented.  !

lit 16- 7 B Process control sheet does not list revision for referenced i drawings.

11 16-8 B Pressurizer code nameplate does not indicate code class.

T11 16-9 B NDE reports do not indicate the 1 governing code. )

11 16-10 B Radiographic flim does not fully comply with ASME Code requirements.

i 1

, A.~ Level I - Violation of licensing commitments, project procedures, or

! engineering requirements with indication of safety concern.

Level 11 - violation of licensing commitments or engineering requirements with no safety concerns i.evel III - Violation of project procedures with no nafety concerns to . Ca t e ryo r y n Hntdware erw'e r t u b Paperwotk cuhte:n C - Program concern

i I

i Table 6.2- 2 RPV HEAD ASSEMBLY ASSESSMENT ITEMS

1. Unit 1 RPV head penetrations assessed by software review.

l 1 Penetration No. Attachment l } CR3M - 03 i 6 CRDM -- 31

4) CRDM -- 11 50 CRDM - 36 56 CROM - 39 63 CLH - 02 I 12 RVLIS adapter l 2. Unit 2 RPV head penetrations seal welded to CRDM housings assessed by visual examination.

i Penetration No. Penetretion,No.

50 60 30 53

54 36 i 61 59 l

!* 73 35 I i 49 56 l 41 42 3

72 66

+

4

)

1 i

i

)

f f

i i

i j

i i

1 0028p/2

l l

l

!/ Table 6.2-3 NONDESTRUCTIVE EXAMINATION ASSESSMENT SAMPl.E PT_ Reports Reviewed PT- 1 PT-22 PT-- 72 PT- 205 PT- 2 PT-29 PT- 76 PT -214 PT- 7 PT-51 PT- 79 PT-230 PT- 8 PT-54 PT-133 PT-251 PT-ll PT- 61 PT-144 PT-260 l

('

\

PT-12 PT-17 PT-67 PT- 69 PT-150 PT- 158 PT-267 PT-19 PT-70 PT-189 PT Examination Witnessed PT-295 Ml' Reports Reviewed ,

MT-3 MT-6 MT- 9 MT-15 MT-4 MT-7 MT-13 MT- 16

() MT-5 MT-8 MT-14 i

i l

RT Reports Reviewed (Weld numbers) l I

BMIA-60 BM1 A- 117 BMIA-139 BMIA-233 i BMIA 6021 BMIA-ll7R1 BMI A- 140 BM] A- 233R1 MAIA-61 BMIA-118 BMI A- 143 BMIA-23322 BMIA- 61R] BMIA- 119RI BMI A- 147 BMIA-233R3 BMIA-68 BMIA-122 BMI A- 148 BMIA-247 BMIA-72 BMIA-126 BMIA-225 BMIA-247R1 BMIA-108 BMI A- 129 BMIA-225R1 BMIA-250 t BMIA- 108R L BM1A- 13321 BMIA-228 BMIA-250R1

! BMIA-110 BMIA- 136R1 BMIA-228R1 l

l O

0028p/3

i l

i 1

EQUIPMENT IN-PROCESS )

INSTALLATION ASSESSMENT CHECKLIST l REACTOR COOLANT PUMP MOTOR ,

l

1. Equipment: I.D./ Tag No.

De sc r ip t ion Serial No. ,

l

2. Receiving indpection report r e f e r e nc ed on installation proces s cont rol sheet . l

3. Inst allat ion process cont rol s heet attributes 2A through 2H complet ed and signed by QC as applicable (preping, l r igg i ng , transporting).

4. Mo t or seating flange cleaned and debur red.

5. Motor leveled to +.125 inch across seating fl arrie d iame t er .

6. Motor lowered into proper cell in manner to preclude damage t o mo t or a rx1 o the r pe r ma ner.t plant equipment.

7 Mo t or set on pump in proper orient at ion, mount ing bolts cleaned, lubed and installed snug tight.

9. Process shee t a nd installat ion Mtivit ies completed in sequence and appli ca.1 e ho ld po i nt s signed by O.C.

9 EMSL requirements imposed with proper time af ter receipt on site.

10. R em ar k s l I

l O RN Team Member Date Page 1 of 1 O

rigure 6.2-1 Equipment In-Process Installation Assessment Checklist for the Reactor Coolant Pump Motor

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l NSSS SUPPORT CIIECKLIST 1

S up por t Id ent i fi cat ion l Drawings Utilized Referenced Dwg. No. Rev. Date j On Process Cont rol Sheet .

u e

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l l L. Gene ra l Assessment 1

i o Configura tion a nd Orient at ion n

o Mat e ria l Ident i ficat ion o Cl ear a nc e s o Freedom f rom obvious damage i

2. Bolt ed a nd Pinned Connect ions I

j o S pacer s i nst a lled (when required) i j o Locking devices f

1 o Proper thread e ngagene nt 1

( o Proper torgaing documented verificat ion 5

3. Field Welded Connections are acceptable wi th j respect to s j o Size j o Contour and Shape o Freedom f rom cracks i

l o overlap J

h' o Slag

.i o S pa t t e r a

j o Porosity i

J, o Arc Strikes 1 o Und ercut i

j page 1 of 2 2

1 0056p/2 l

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4 Figure 6.2-2 NSSS Support Checklist (Sheet 1 of 2)

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4. Process Control Sheets 4

i o Appropriat e ent rias/signat ures J

j o Hold point s signed by qua lif t eri i nd i v i d uc i a 1

1

] o Procedure and drawing revisions correct for j t i me of installation.

t j o We ld er s were qualified i

5. P rocu r em e nt Doc ume n t a t ion 1

l o Weld f i l t er ma teri al acceptable 1 Type HT. No. Size j Type HT. No. Size {

j o Documentation required by the pO/ Spec J is available

- Cod e d at a reports compl e t e

'4 6. Deviation Reports are properly prepared, have

-1 the appropri at e d ispostion approva l signatures, i and properly closed (when applicable)

I

.; DR No.

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. Team Member Date j

i page 2 of 2 l 0056p/3 1

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i Figure 6.2-2 NSSS Support Checklist (Sheet 2 of 2) 9

$ . . . , - .- , , , , , . ~ , . - - . . - - , , , , ,w,,_w,

_ _ - - _ ~ - - - -- .- - - - - - . - - - - - - - - - - - - - - - - - - _ _ .

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MODUL E 10 - CONSTRUCTION ASSESS?1CNT CHJCKLIST -

} MSSS MECHANICAL EQUIPMC4T i

j i

j STEAM CENERATOR

] SPIN Number 1

i ASME Code Class J

1 ASME S/N Tag Number 1

1 i

i j Field Assessment i

j 1. Check for equipment id ent i fi ca t ion.

i

] 2. Equ i pme n t layout is correct in relationship to p r e ac t or loops.

Installation Documentation l

3. Check the setting proc ess cont rol s he e t fors o Appropriate entries / signatures o liold points signed 4 Check p roced ure revisions listed on the process cont rol sheet are cor rect for the t ime of installation / inspection.

5. Check to a ssure the spe c i f i ca t ion comm i t me r.t s listed below have been adequately addressed a r:1 i n spe c t ed by NI SCO d ur i ng t he installat ion /

setting process.

o In spe c t a nd clean st eam genera tor pri or t o rough placement in preparation for final set:ing, o Check the st eam genera tor a l ignme nt so that the generator in the vertical position is within the t oler a nc e o f + 0 d eg r e e n 30 minutes (1/2*) vertically. The steam shall be set t o the predetermined opt in:m eleva t ion t o wi thin _+ 1/3 2 inch.

0056p/4 Figure 6.2-3 Construction Assessment Checklist for the Steam Generator (Sheet 1 of 2) n .e

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J j Procurement Doc um e n t a t ion 1 6. Quality Ass urance documentatici, required by the ~

i P.O./ Spec is available.

t i

j 7 Cod e d a t a report f orms avail *ble and comp;et a, o Report references cor rect s er i a l n umbe r a nd cod e cla s s .

1 e Certificat e of shop inspect ion complete.

complete.

! Ge nera 1 Docume nt at ion 1

j H. Deviation reports are properly prepared, have the

appropriate disposition approval signatures, a nd

properly closed (when applicable) .

1 1 DR No.

4 j DR No.

j DR No.

a 1

9 D ra wi ng revisions referenced on the proceas i cont rol sheet are correct for the time of i i n sta 11at t on/in spec t ion.

! O o

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i 10. EMS L requ ir e me nt s i mposed wi thin proper t ime j after receipt on site.

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Team Member Date: -

i Page 2 of 2 i 0056p/5 j 1 l 1

i l l Figure 6.2-3 Construction Assessment Checklist for the j Steam Generator (Sheet 2 of 2)

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_ . - . . _ . . . _ - , - . . , , , , , . . - .--m , , . , - . r-

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1 MODULE 16 - CONSTRUCTION ASSESSMENT CHECKLIST .

NSSS MECHANICAL EQUIPMENT PRESSURIZER SPIN Number ASME Code Class ASME S/N Tag Number Field Assessment

1. Check for equipment ident i f i ca t ica.

2. Equipment layout is correct in relationship to r e ac t or loops.

Installation Documentation I

3. Check the setting process control sheet fort o Appropriat e ent ries / signatures o ho ld po i nt s signed

4. Check procedure revisions li sted on the prc ess cont ro l sheet are cor re c t for the t ime of installation / inspection.

5. Che ck t o a s s ur e the spe c i f i ca t ion c ommi t ne r. a listed below have been adequately addressec and inspected by NISCO during the i n st a llat ion /

setting process, o In spec t and clean pressurizer vessel prior to rough placement in preparation for final setting.

o Lower the pressurizer vessel t o approx ima t ely one foot above the foundation. Examine the vessel orient at ion a rd ma t ching wi th bolt holes. The vessel suppor t ring should be inet slied at thi s t ine.

Page 1 of 2 0956p/6 Figure 6.2-4 Construction Assessment Checklist for the Pressurizer (Sheet 1 of 2)

- -. ,. . ._o-.--,.,,-, ..--...,___w. --n-._.._m.y-___.

9

5. (cont'd) o Lower the pressurizer to the foundation and install the hold down bolts. Verify that the vertical position is within 2 degrees 30 minutes (1/2*) of vertica l. The pressurizer shall be set to the predetermined optimum elevation to within 31/4 inch.

o Install the upper lat eral supports around the pressurizer lugs making sure that the slots which fit around the pressurizer ILgs are shimme 2 to the required cold gap cl e a ra nce.

o Torque the hold down bolt to provide a bolt tension of 186 min, 206 max, KIPS.

P rocur e me nt Docu me nt a t i on

6. Quality Assurance document at ion required by the P.O./ Spec is available.

7 Code data report forms available a nd complete.

o Report ref erence s correct serial number ard code class, o Cert ifica t e of shop inspection complete.

General Documentation

8. Deviation reports are properly prepared, have the appropriate disposition approval signatures, a nd properly closed (when applicable) .

DR No.

DR No.

}. DR No.

1 l 9. Drawing revisions referenced on the process )

j cont rol sheet are correct for the t ime of i n st a llat ion / i n spec t ion.

i i o Rev i o Rev

] o Rev l o Rev i

Team Member Dater j 0056p/7 4

O Figure 6.2-4 Construction Assessment Checklist for

, the Pressurizer (Sheet 2 of 2) s j

5

1 1

% j i

PROCESS CONTROL SHEET <

1 j ASSESSMENT CHECKLIST

'l j 1. Process Control Sheet No.

/ Item: Description Serial No.

l ASME Class e

d

2. Process Cont rol Sheet Hold Points:

I l Coupl e t ed i Signed by Certi fied Personnel 1

l j 3. Applica ble DRs Re f e r e nc ed Reviewed 4

DR Nos 1

i 3 4. Ent ri es l eg ib le a rd a ny cha nge s we r e made properly l

l S. Spec i f icat io n requirements are properly reflected

a rri litpl e me nt ed

6. Procesa Control Sheet i s admini st rat ively ar.d

]

3 t echni ca 11y ac cur a t e

, 7. Support ing documents are attached or referenced as applicable (check for 6 CRDMs/CCHs)

8. Welders / welding operat ors quali fied (check for

welds per formed on 6 penetrat ions) 4 i 9. Remarks

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j RR Team Member Date 4

1 0056p/8

.s Figure 6.2-5 Process Control Sheet Assessment Checklist i

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React or Internals Assembiv a nd Installation Checklist Process Cont rol Sheets Reviewed 1 I )

1 l 2 l

- 1 I 1

Drawings Utilized Referenced Dwo. No. Rev. Date i on Process Control Sheet ... )

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1. Process cont rol sheet

] o Appropriate entries /signat ures e

h o Ibid points signed by qualified i nd i viduals

. o Procedure a nd drawing revi sions cor rect for j t ime of installation t

l 2. Procur ene nt d ocu me nt at ion d

1 o Documentation required by the P.O./ spec is l available

\

l 3. Deviation reports are properly prepared, have

! the appropriate disposition approval signatures, and properly closed (when applicable) .

! DR No.

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t 4 4. The following specifi ca t ion requirements have been adequately address-d by the inst allat ion docume nts ,

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s Page 1 of 5 j 0056p/9 i

f Figure 6.2-6 Reactor Internals Assembly and Installation

Checklist (Sheet 1 of 5) t 1 l

G O 4. (cont'd) Requirement implementation Location o Prior to initiating assembly, all accessible recctor internals parts and welds shall bevisually

9 exami ned f or def ects . The fult length of all major girth welds sha ll be e xa mi ned. -

o Clean all components thoroughly before a rd a f t er a s s e mb ly in accordance wi th We s t i ng hou s e Process Speci ficat ion PS 597760. Maintain clean room conditions d uring assembly ___

of i nt er na l s .

o Activities requiring quality assurance _

check, veri fi cat ion, or witnessing by We st inghouse per sonnel mu st be i dent if i ed l and coordinated with appropriat e parties l

prior t o i ni t iat ing wor k.

o Before lif t ing components, lock tab welds on the int ernals lif ting rig shall be checked. Roto-Lock inserts which are used fo; attaching the lift rig t o the internals shall be checked to insure they are inst alled a s ma tch-ma rked pa irs in identical locations in the core barrel ard the t r pe r inter nals s upport plate.

Correct oriantation of steis on the lif ting rig shall be verified. Insert-to-flange linws must be aligned.

o ilachine the RV radial s upport blocks in accorda nce wi th the West inghouse lower railial s qpport clevis machining details.

o In s pec t for proper perpendicularity of support blocks, tapped holes, ard flatness of pads on the lower radial s up po r t clevises.

w, o Lift lower internals assembly clear of s t or a g e s t a nd . The a ssembly must be level wi thin 0.10 inche e total . Record the we ight on the l oad s er.sor s when the assembly is f reely suspended.

o Measure a nd record the f riction load weight reduction as the lower radial sgppggt ggg,3ngages the clev! s members 0056p/10 O

rigure 6.2-6 Heactor Internals Assembly and I n n t .a l l a t. i o n Checklist (Sheet 2 of 5)

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4. (cont'd) Implementation i Location I

o A radial cleara nce of approximately

. -0.080 inches should exist at matching

a nozzle int er f ace s . At each nozzle, at i four equally spaced locations, measure

<t and record the clearance gap between the b ar re l a nd ve s sel . Also measure and record the minimum gap at each '

, nozzle.

1 o Take measureme nts a nd machine the l o we r radial support elevis. Requirements are i s ho wn o n We st inghou s e ga ug i ng a nd a s s e mb ly document s for the lower radial support c'. evi s i n s er t . Thi s i s a second machining operation.

o Machine the lower support clevis inserts g in accordance with the Westinghouse r equ ir e me nt s .

5 o Torque bolts on the head and vessel

, a lignme nt pins in accordance with  !

• s Westinghouse requirements.

4 l o Install the hold-down spring on the upper core barrel flange. Take a nd record necessary measuriments in accordance with We st inghouse inst ruct ions.

o In s pec t the lower radial support c l e vi s j i nser t i p ri or to and af ter welding l ock bars and dowel pins. Verify that the

• lower radial s upport elevi s inserts are i are seated in the RV and that no gaps exist on s id e a nd t op. If gaps exist, stop work and not if y West inghouse.

1 o Me a sur e a nd record the cleara nce, t op -

and bottom, at each of the lower radial s up por t ke ys .

4. sc ~

o A minimum clear space of 48 inches below the upper internals assembly for access ___

, to fuel assenbly guide pins, for checking thermocouple insertion a nd, for

inspection of the guide tube profile.

o A minimum clear space of 120 inche s a,bove the upper internals assembly for

,. the r mocoupl e in s er t i on .

P6ge 3 of 5 4

0956p/11 4

Figure 6.2-6 Reactor Internals Assembly and Installation

} Checklist (Sheet 3 of 5) 4

't

. ~- g -.w-, ,>v- ,- - -,---,,y- +yg em -

O c

4. ( co n t ' d ) Impi eme n t a t io n Location a Place in the assembly stand. Do not re no ve the protective covering u nt il ready to set the upper internals assembly into the core bar re l (in the RV).

o Level upper internals assembly flange to w i th i n .0005 inches per foot.

o The apper surface of the core bar rel mu st be l eve l t o w i thi n 0.18 i nche s.

Record flange the weight indicated on the load s ensor .

o Apply Neolube to the upper core plat e aligning pins ard the upper core plate inserts. Allow to dry.

o I n spec t and verif y pin-t o-pin spac ing of the f ue l pins located around the periphery of the upper core plat e.

o Che ck a nd record clearance for the f uel a ssembly all arourd the baffle at the upper and lower core plate. Use the We st inghouse ga uge s for fuel a ssembly guide pin and baf fle checking, o Ver i f y proper spacing of f uel a ssembly guide pins on the upper and lower core plates. Use the We st inghouse ga uges for fuel assembly guide pin and baffle checking.

o Meas ure and record the radial clearance between the up pe r s up po r t fl a ng e a nd the inside-diamet er of the vessel above the support ledge. _

o Meas ure and record the radial cle a ra nce between the core barrel fla ng e a nd the inside-diamet er of the RV above the s up po r t ledge.

o Measure and record the gaps between the upper core plate inserts a rd the upper core plate guide pins. Compare results to those recorded on as-built drawings. Subni t results to We st inghouse f or review.

Page 4 of 5 C

Figure 6.2-6 Reactor Internals Assembly and Installation Checklist (Sheet 4 of 5)

I

4. (cont'd) Implementation Location a Mea s u r e a nd record the gaps between the uple r support fl a nge a nd the head a nd vessel alignment pins. Compare results to those recorded on as-built d ra wi ng s .

Su bmi t results t o We st inghouse for review.

o Me a s ur e and record the gaps between the upper core plate and core barrel band ad jac e nt to the 0*, 180*, 90*, a nd 270*

a xe s a s wel l a s positions 45' r em oved ,

from the axes. Submit result to westinghouse for review.

o Resolve all West inghouse review comment s .

o Verify that the core barrel fla nge is level within .010 inches t ota l .

o Verify that all mating surfaces are f u lly seated a nd clamped. We ld the locking devices.

o Ver i f y, by ha nd, that there is no looseness of locking collars on the i n st ru me nt guide columns beneath the lower core plate. Install the locking d owe ls .

o verif y final torquing of the locking d owe ls o f bo t tom mo unt ed i n st rume nt at ion and that there is a free path for flux t h i nb l es ,

o Verify that the flow nozzles have been installed a s required.

O Te am Member Date:

0056p/13 Pigure 6.2-6 Reactor Internals Assembly and Installation Checklist (Sheet 5 of 5) e

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MODULE 16 CONSTRUCTION TEAM CHECKLIST - BMI GUIDE TUBES I l

Guide Tube Number J

d Drawing Number l

1

1. Guide t ube is free f rom sur f ace damage . . .

2. Tubing is installed per drawing . . . . . ..... . _

f t

3. Socket welds - visual examination are acceptable for

{ the following attr ibutes:

1 t

j Un6arcut . . . . . . . . . .... .

Suface defects . . . . . . . . . . . .

We ld contour . . . . . . . . .

3 1 Wald size . . . . . . . . . . . . . . . . ..

4 Note: Acc ep ta nc e c r i t er i a pe r NI CCO P rocedur e ES-100-5, visual inspection of welds.

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0056p/14 i

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j Piqure 6.2-7 Construction Team Checklist for BMI t

Guide Tubes (Sheet 1 of 2)

- - - - . . . - , , - , - - - . . - _......-.--...--m.., - . . . , ,.~ v ---.~.--.-

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j MODULE 16 4 CONSTRUCTION TEAM CHECKLIST - BMI GUIDE TUBES 1

l. All hold points, CA a nd ANI are appropriately

{ signed. .

I l 2. P" reports f or socke t weld root passe s and j

final welds are at tached t o the process control sheet. .

s l 3. For three d if ferent GA ho ld po i nt s , the procedure i revision listed on the process cont rol sheet is i car re c t for the time of inspection.

1 Procedure /Rev. Date of Inspection

4. For two different GA hold points, the inspector s were certified at the t ime of inspection j In spect or Dat e of inspect ion t

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( b ac k )

0056p/15 rigure 6.2-7 Construction Team Checklist for BMI l

Guide Tubes (Sheet 2 of 2)

MODUT.E 16 CONSTRUCTICN TFJJi CHECKL!ST - DMI GUIDE TULE SUPPORTS

1. For (2) pieces of mat erial - the identification, typa, and size is as required by the 3 r a wi ng s Dwg No It em No flea t No Dwg No Item No lle a t No .

2. For the above (2) pieces of mat erial - the C.'1TR ' s are in enmp l i a nc e with code requ ir e me nt s . .

3. For (5) piece s of material - the mat eria l identit y is ev i de nt a rr3 if required, transf erred prict t o cu t t ing .

Dwg No Item No . .

Dwg No Item No .

Dwg No It em No Dwg No Item No Dwg No It em No .

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page 1 of 2 l f r o nt )

0056p/16 O n l'iqure 6. 2-8 Construction Team Checklist for BMI Guide Tube Supports (Eheet 1 of 2)

fr l

MODULE 16 CONSTRUCTION TEAM CHECKLIST - BMI GUIDE TU3E SUPPOPTS

4. For (5) welds - visual examination is acceptable for the f ollowing at tribut es:

?

Und er cut . . . . . . . . . . _

Sur f ace damage . . . . . . .

Cont our . . . . . . . . . .

Size . . . . . . . . . . . .

z Note: Marked-up drawings (s) i ll ustra t i ng the welds as ses sed i s a t.tached. Acceptance cr;teria per NISCO procedure E. S . - 10 0- 5, visual inspection of welds.

Team Member Da t e :

page 2 of 2

( back )

0056p/17 Pigure 6.2-8 Construction Team Checklist for BMI Guide Tube Supports (Sneet 2 of 2)

1 I

i MODULE 16 CONSTRUCTION TEAM CHECKLIST - RECEIVING INS P E.CTION REPORTS NISCO RIR NUMBER l

1 L. The RIR ident if ies the mat er ial/ component _ l Identity / Tag Numbers Heat / Lot / Serial Numbers Size / Type

2. *he ma ter ial cert if ica t ions, data reports ard piping mat erial s requisitions. as applicable, are available and/or at tached. . . . . . .

3. The Col C, CMTRs and data reports, as applicable, are acceptable for the following at tribut es: .

Leg i bi li t y . . . . . . . . . . . . . . .

Cer t ificat ion by signat ure of the s up pl l er o r authorized re sp re se ntat ive . . . . . . .

l Compl i a nc e w i th the requireme nts of the code a nd .

l specifications. . . . . . . . .

Traceabi lity to t he item received .

4. The RIR snows appropriate sugnatures by CA/CC for all l inspections performed . . . . .

5. The RIR shows acceptance or rejection of the _

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> Team Metrber Date:

OOS6p/lB I

Fiqure 6.2-9 Construction Team Checklist for Receiving Inspection Reports

l

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1 MODULE 16 CONSTRUCTION TEAM CHECKLIST - WELDING MATERIAL STORAGC Storage Location Date Surveyed

1. Acces s cont rol s are in place . . . . . . . . . ...

2. Hou sek e epi ng a nd t enpe r a t ur e are within requ ir eme nt s ..

3. Ident i fication of filler material is evident . . ...

4. Low hydrogen electrodes stored per requirement . . ..

S. Holding oven temperature and calibration is appropriate.

6 Holding oven cont ent s segrega t ed by electrode type and size . . . . . . . . . . . . . . . . . . . . ...

Team Member Date:

0056p/19 Figure 6.2-10 Construction Team Checklist for Welding Material Storage i

I IF MODULE 16 CONSTRUCTION TEAM CHCCKLIST . WELD MATERIAL CONTROL

/

DATE OF SURVEY W EID ER WELD NINBER

1. Weld material cont ainer has a properly completed weld material requisition . . . . . . . . . . . . . . . _

2. Moisture sensitive electrodes are i s s ued / st or e d in heated containers ........... . . . .

3. Por table rod ovens are holding mi nimum t emperat ure . . _

4. Weld material containers have only one type or classification of weld filler ma t erial . . . . . . . . _

5. The completed WMR shows evidence of return of stubs /

unused weld mat er ial . . . . . . . . . . . . . . . _

6. The completed WMR has been placed in the appropriate J. rocess cont rol sheet file . . . . . . . . . . . .

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0056p/ 20 rigure 6.2-31 Construction Team Checklist for Weld Material Control

)

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MODULE 16 - CONSTRUCTION l

FIELD ASSESSMENT CHECKLIST - LIOUID P ENI T R ANT I

i REPORT NO:

1. Cor:pa r e the practical application of the P.T. e xa mi n a t ion performed by the NISCO inspector with the NI5CO P. T.

procedur e requirements to insure the following parame ter s are mett o Exami ner Qualificat ion Level o Consumable Mat erials o Te mpe r a t ur e Ra ng e o Cleanilness of Sur f ace to be Examined o Pe ne t r a nt ( Applying ard Re mova l )

o De ve l ope r ( Applying and Removal) o Evaluation of Ind i ca t ions o Hold / Dry Times 2 Review the liquid penetrant inspect ion report to insure the i

following paramet ers are met l

o Legibi li t y o App r op ri a t e Ent rie s / Signat u re s o Admi ni st ra t iv e Completeness e

O DATE 9 RE AD I N E.SS REVI EW TEAM M EMB ER 0056p/21 0 -

Figure 6.2-12 Field Assessments Checklist for Liquid Penetrant e

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, MODULE 16 - CONSTRUCTION SOFTWARE REVI L'W CHECKLIST - LIQUID PEh5TRANT REPORT NO:

(

l 1. Compa r e the liquid penetra nt inspection report information ,

f with the NDE proced ure t o insure the following paramet ers l J are met: 1 l o Test Part Ident if s cat ion 4

j o NDE Procedure No./Rev. / I 1

o Dwg. No., Ma t er i a l . Material Thickness 3

o Cons umable Mat erials  ;

j o Exa ni nat ion Da t e 4

o and Qualification Level of the l

a ier ard Ac cep t or

! I j o Examination Results f o .roject Number 4

i j 2. Assess the completed l iquid penet rant inspect ion repor t to j a scer ta in whether the following at tributes are proper i

o Legibility j o Appropriate Entries o Administrative Completeness

)

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j RE AD IN ES S REVI EW T EAM M DM B ER DATE 1

. 0056p/22 4

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, Figure 6.2-13 Software Review Checklist for Liquid i Penetrant i

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t 1 MODULE 16 - CONSTRUCTION 3

l SOFTWARE REVIEW CHECKLIST - MAGNETIC PARTICLE t

f REPORT NUMBER:

i l

i 1. compa re the magnetic particle inspection report information with the NDE procedure to insure the following paramet ers

} are mets 1 i o NDE Proc ed ur e No. / Rev . /

j o Test Part Identification o Consumable Mat erials o Examination Date i I o Project Number i l

o Governing Code j- o Pertinent Technique Data / Type of (Magnetization, 1 Current Density, Prod. Spacing, Coil l

} Configurat ion etc. ) l 1

{ o Dwg. No., Mat e ri a l , Material Thickness o Na rap ** Qualification Level of the Exar.iner a nd c.. cept or

( o Exa mi nat ion Result s i

2. Asses s the completed magnetic part icle s inspection report to a scer tain whether the following attributes are proper:  ;

i o Legibility

.j o Appropriat e Ent ries i

. o Admini st rat ive Completeness I

]

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i j READ IN ES S REVI EW T EAM M EM B ER DATE 1

.J 0056p/23 4

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i l Figure 6.2-14 Software Review Checklist for Magnetic j Particle j

s 4

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) Module 16 - Construction j

Software Revi ew Check l ist - Rad iog raphy i REPORT NUMBER 1

i Weld / Support Number SS CS j

j Drawing Number 4 1 t 1. Compare reader sheet infornation with ND3 procedure to insure the .

f o llowi ng pa r ame t er s a re m.it s o ??CE procedure No./Rev /

o Cassettes and lead screens used are acceptable o Pene t ra me t er size, t ype , l oca t ion, a nd s hi mmi ng f

! o Source size i o Material thickness l o Source to film distance

! o We ld ident ification ident ified on the ra d iog r aphy film agree with d radiographic examination report I

i o Level II or III int erpre ter qualifica t ions are j current o ANI film review acceptable

+

2. Revi ew radi og raph s for the following attributes:

)

{ o Proper storage 1

o Proper film type o Film ID markers are cor rect, properly loca ttd, a nd legible

)

l o Film density acceptable 1

{ o Film art if acts q

l i o Int erpretat ion agrees with reader sheet l

j i

4 i

j Feadiness Review Team Member Lev e l Date l oc56p/24 1

1

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1 J

Figure 6.2-15 Software Review Checklist for Radiography i

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/"N 7.0 INDEPENDENT DESIGN REVIEW The Readiness Review process includes an independent review of design documents, such as system design criteria, calculations, specifications, and drawings, to ascertain whether proper design requirements were considered and that design documents correctly implement licensing commitments. This review is being conducted

/~%)

(_ by Stone and Webster Engineering Corporation and includes a team of technical and professional experts to assess the technical adequacy of the design of the work covered by this and other modules. This independent design review (IDR) ensures that technically the design meets commitments, and that the

("'

mechanical equipment, piping, and components within the scope of this module are in fact adequate. The results of the independent design review will be provided as a separate report.

The IDR is conducted in two parts. One is an integrated evaluation of selected systems across all disciplines: the second is an evaluation of the structural aspects [ reinforced concrete (Module 1), structural steel (Module 8), foundation materials and backfill (Module 13A), etc.] of the project design. The integrated IDR (including nuclear steam supply system) will be issued as a separate report, whereas the structural reviews will be issued with the. appropriate structural module.

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O O 0068p/324-5

8.0 PROGRAM ASSESSMENT /C0NCLUSION O

8.1

SUMMARY

OF OPEN CORRECTIVE ACTIONS 8.1.1 SECTION 6.1 o Finding 16-11 Action: Revise cons t ruction procedure GD- T- 01 to clarify the scope of Deviation Reports (DR).

Responsible Organization: Construction Completion Date: November 15, 1985 Action: Review all past CFR 50.55(e) reports to determine status of corrective actions and add open actions to the commitment tracking log.

Responsible Organization: Construction i Completion Date: November 26, 1935 o Finding 16-12

_ Action: Include actual versus allowable acceleration comparison in stress calculations during routine l

as built reconciliation. i i

l Responsible Organization: Bechtel Engineering 1 Completion Date: Completion of As-Built reconciliation o Finding 16-13 Action: Issue desk instruction and provide training to institute a program that vendor drawings be reviewed and cross referenced on instrument installation drawings to l( assure that vendor requirements are included in the installation drawings.

Responsible Organization: Bechtel PFE l g-w Completion Date: November 22, 1985 l(

l O

4 o Finding 16-15 l

Action: Revise Project Reference Manual, Part C. I Section 37, to include details of the program for verification of Westinghouse equipment seismic

, qualifications.

Responsible Organization

Bechtel Engineering Completion Date: December 2, 1985 j

8.1.2 SEC'110N 6. 2 1

There were no open actions remaining from Section 6.2.

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O 0073p/324-5 0

8.1-2

I 8.2 QA STATEMENT The process for the development of this mcdule was monitored by the Readiness Review Quality Assurance (QA) staff for general adequacy.

The primary focus of the monitoring effort. was the

, 's f~)T identification, documentation, analysis, and resolution of Readiness Review Findings. The finding reports issued by the Readiness Review Team and their responses were reviewed, individually and collectively, for root causes and generic issues; i.e., trends. Based upon review of the responses and commitments to individual finding reports and generic concerns, the resolutions were determined to be adequate. All findings j were initially distributed to project QA for review for reportability [10 CFR 21, 10 CFR 50.55(e).1 in accordance with existing QA procedures. In addition, findings were screened by Headiness Review to determine whether any required additional evaluation by the project for reportability. One was so identified, but was subsequently determined to be nonreportable by the project.

Other monitoring activities consisted of reviewing personnel qualification and training records for the team members, reviewing the verification plan, and reviewing completed checklists to assure adequate identification of findings.

(f--) Additionally, an independent reverification was performed on a sampling basis under Readiness Review QA overview to determine the adequacy of the Commitment /Implemer.tation Matrixes and the Design / Construction verification efforts.

Based upon these monitoring efforts, this appendix and the Readiness Review Team conclusions are judged to be acceptable.

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<Deor(e C. Bell 'lohn II.

. Draqqn_._-

Readiness Review Team Headineau Heview Tediu Quality Assurance Representative Quality Assurance Representative 0071p/323-5

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G MODULE 16 TECHNICAL CONSULTANTS CERTITICATION On the basis of review of the Module 16 report on NSSS interface and NSSS construction, the project organization and selected documents, I certify that to the best of my belief and D knowledge the information and conclusions contained herein are factually and technically correct. Under the program described in Section 4 and the corrective action described in Section 6, the commitments of the VEGP FSAR are being implemented and the NSSS interfaces and construction comply with FSAR commitments.

( V C f!

Russell P. Bone i

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I Nuclear Steam Supply System - Module 16 Readiness Review Board Acceptance

! The Readiness Review Board has been apprised of the scope and content of Module 16, Nuclear Steam Supply System.

The Board has reviewed the program verification, as well as corrective actions, both proposed and implemented, by the Vogtle Project. Based

,P upon this review and based upon the collective experience and profess-ional judgment of the members, the Readiness Review Board is of the opinion that the corrective actions are acceptable, and that the nuclear i steam supply system at Plant Vogtle is sound and con. plies with commit-ments set forth in the FSAR and acceptable practices.

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w r h h APPROVED: __

h_ DATE: Ic 5

_ Dutton T Doug Chairman, Readiness Review Board

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Vogtle Electric Generating Plant 1

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. . . .. . . _ . - - - . . - . . - . ~ . . . . . . . . - - . . _ . . . - . ,.. .. . _ . .

i l

Georgia Power Company Project Management ,

l Post Office Box 282 [

Waynesboro, Georgia 30830

- Telephone 404 724 8114 404 554 9961 I i Southern Company Services, Inc.

Post Ottice Box 2625 .

4 Birmingham, Alabama 35202 Telephone 205 870-6011 gle El i

i

! : Date: November 19, 1985  !

Re: Plant Vogtle - Units 1 & 2 Readiness Review Mcdule 16

! File: X7BD102 -

< Log: SS-5403 From: 0. Batum t

T.o : W. C. Ramsey Engineering has reviewed Module 16, Nuclear Steam Supply System,for general accuracy and completeness. To the best of our knowledge and belief, the module is a complete and accurate representation of the l l Nuclear Steam Supply System, and the engineering process and commitments related thereto.

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LU41 0

! Ozen Batum General Manager, Project Engineering - Vogtle

( xc: Project File

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m a.m . ..-..m Geor; a Power Company Pro,ect Manageme".t Route 2. Box 299A WaynesDoro, Georg a 30630 Teiepnone 404 724-8114 0 404 554 9961 L

Vogtle Project O

DATE: August 21, 1985 RE: Plant Voctle - Units 1 & 2 Readiness Review Module 16 FROM: M. H. Googe TO: W. C. Ramsey Nuclear Construction has reviewed Module 16 excluding the referenced appendices. To the best of our knowledge and belief, the module is a complete and accurate representation of the Nuclear Steam Supply System Installation Program and commitments related thereto.

\

/ .ALANh ,5?'~t M. H. Googe / //

Project Consttuction Manager II Vogtle Nuclear Construction Department O

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_ . . . -. . - - .- - . .~

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l 8.6 RESUMES

( .

The following resumes, arranged in alphabetical order .present a  !

brief professional history of those people instrumental to the development of Module 16. .

ROBERT D. ANDREWS, Piping Engineer,. Design Team Member Mr. Andrews has been with Bechtel for over 18 years, fulfilling  !

various assignments associated with the design of mechanical {

systems in. mills, smelters, and refineries.before being assigned l l to supervise the verification of piping systems and penetration  !

seals in Unit 2 of the Diablo Canyon nuclear power' plant. He -

also supervised the layout of piping and equipment for facilities being added to the plant.  ;

Before he-joined Bechtel, Mr. Andraws spent 16 years, working on h

-the design of steel mill equipment.

?

Education: l University of California ,

B.S., Mechanical Engineering f- P.E., State of California GEORGE C. BELL, Quality Assurance Engineer, Quality Assurance Representative  !

Mr. Bell has over 12 years of experience in the nuclear power plant industry, 8 of them in Quality Assurance. Most recently "

! he was assigned to the Limerick Nuclear Generating Station as l

project quality assurance engineer. His principal responsibilities included: coordinating the preparation of and establishing the Quality Assurance Program within the' Project including the Operational Quality Assurance Program for Bechtel work inside the operation boundaries of Unit 1:-directing quality activities on the project and reviewing project schedules for quality-related activities to ensure timely and effective implementation of the Project Quality Assurance Program; and supervision and providing work assignments to Project Quality Assurance personnel (site and' home of fice) and preparing reports on the effectiveness of Quality Assurance Program implementation. Previous assignments have been the O- .

W. H. Zimmer Nuclear Station, Susquehanna Steam Electric Station, and Davis-Besse Nuclear Power Project.

Education:

Villanova University O B.E., Electrical Engineering

M.B.A., Wilkes College of Pennsylvania P.E., State of Pennsylvania l

RUSSELL P. BONE, Supervisor Advisory Operations Division, i

Technical Expert Mr. Bone joined Stone & Webster Engineering Corporation in April  !

1971, and is presently serving as Supervisor of the Operations  !

Design Review Group in the Advisory Operations Division. Mr. l Bone has over 14 years experience in the e";ineering, design, '

start- up testing, and operation of power plants. He is a

{

Mechanical Engineer with additional education in nuclear I engineering and experience in the operation and design of nuclear and fossil power plants.

Prior to joining Stone & Webster, he served as Chief of Quality Assurance for the U.S. Army Nuclear Power Program, where he i qualified as Officer in Charge of an Army Nuclear Power Plant. )

Education:

l l

Worcester Polytechnic Institute B.S., Mechanical Engineering RAINER M. GRIMM, Engineering Contract Administrator, Design Team Member Mr. Grimm has over 20 years experience in the nuclear power industry in Europe, Asia, and the United States. He has served in various capacities on the Rancho Seco and San Onofre projects. Most recently, Mr. Grimm served as administrator for the nuclear steam supply system and turbine generator contracts for power units in South Korea. l Education:

University of Munich Diploma, Chemical Engineering )

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KENNETil J . IIAY , Senior Engineer, Design Team Member i l

Mr. IIay has been assigned to five nuclear power plant projects spanning a period of 14 years. Principally engaged in pipe stress and support analysis, he has also served as ASME Codes Engineer and was responsible engineer for several material procurement c o n t r a c t.s . hl O

8.6-2

i

jg Education:

'\_)

i Kingston College of Technology ,

Univeruity of London, England liigher National Certificate - Mechanical Engineering JAMES H. HUME, Project Manager, Design Team Member i

Prior to joining Readiness Review, Mr. Hunie spent 6 years in project management positions on the Maanshan (Taiwan) and Korea Units 7 and 8 projects. Previously he served as design project engineer on San Onofre and two European generating plants..

( Before joining Bechtel, Mt. Hume was responsible for directing conceptional design efforts on several large reactor projects for Atonica International and also supervised a systems analysis  ;

unit that performed design, analysis, and optimization of reactor systems.

Education:

l Georgia Institute of Technology B.S., Industrial Engineering University of New Mexico M.S., Mechanical-Nuclear Engineering I.MO P. IlYDRICK, JR., Instrumentation Quality Control Engineer, ,

Construction Team Member l Mr. Ilydrick has 7 years of nuclear experience in quality control ,

engineering with Bechtel Power Corporation. His  !

responsibilities have included supervision of inspection I activities required to complete construction of Diablo Canyon Nuclear Units 1 and 2; performing software review activities

)

)

associated with quality assurance records completion for Limerick Nuclear Units 3 and 2; developing the quality control program and procedures required for the completion of construction at Zimmer Nuclear Unit 1; and the supervision of

() inspection activities associated with the construction of Susquehanna Steam Electric Station Nuclear Units 1 and 2.

Education:

i Pennsylvania State University

() Nuclear Engineering Technology l JAMES L. MARTIN, Quality Assurance Engineer, Design Team Member Mr. Martin's 19-year career includes 16 years in the nuclear

() industry. 4 in submarines and 12 in power plants. In addition I

8. 6- 3  !

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to quality assurance engineering assignments at Enrico Fermi, limerick, Oconce, Edwin 1. Hatch, and Grand Gulf, Mr. Martin has had design responsibilities at Davis-Besse and Hatch. During his association with Lie nuclear navy, he designed piping systems and equipment foundations.

Education:

Richland Technical College A.S., Civil Engineering Technology l LESTER J. MIKULECKY, JR., Construction Engineer, Construction Team Member Mr. Mikulecky has 5 years of nuclear power plant experience and h is currently employed with Georgia Power Company. His experience includes writing installation procedures for a quality assurance ASME manual, and serving as a technical specialist for pipe supports, and a system engineer for reactor coolant and utility water system. Prior to joining Georgia i

Power Company he was employed with Daniel International Corporation as a certified quality inspector 11. His responsibilities included visually inspecting safety and non-safety related welds; monitoring qualification of welders; performing component support installation inspections for compliance to applicalle drawings, specifications, codes, and procedure requiremente: and assisting in the preparation of procedures and reportr.

Education:

lowa State University B.S., Construction Engineering ROBERT W. McMANUS, Ast:istant Project Construction Manager, Construction Discipil e Manager Mr. McManus has been with Georgia Power Company for over 11 years, 5 of them on direct assignment at the Vogtle Electric Generating Plant. He was most recently responsible for the quality acceptance of Civil, Electrical, and Mechanical portions h of VEGP. Responsibilities other than management of personnel included reviewing Field Change Notices to design drawings for acceptance, contact with Engineering Quality Assurance on acceptability of the site quality program, construction contact for the Nuclear Regulatory Commission for their quality audits, and performing departmental audits of site construction h activities for design compliance.

Education:

Southern Technical Institute B.S., Civil Engineering Technology lh 8.6-4

CHALMEb R. MYER, Engineering Group Supervisor, Design Team Leader Mr. Myer has been involved in the design and analysis of piping, pipe supports, and mechanical equipment on various nuclear generat ing plant projects for the past 12 years. In his most recent assignment, he supervised the finalization of mechanical

("T equipment systems design and the resolution of construction and

(/ startup problems.

Education:

New Mexico State University U.S., Mechanical Engineering P.E., State of California WILLIAM C. RAMSEY, JR., Readiness Review Program Manager >

Mr. Ramsey is employed by Southern Company Services as manager- nuclear projects, nuclear safety i,nd fuel. He is currently relieved of those position responsibilities and assigned management responsibilities for the Vogtle Readiness Review Program.

lie has approximately 13 years of nuclear experience that encompasses design engineering, engineering management, O. s c hedu li ng , budgeting. procedures administration, testing, ntartup, engineering evaluations and audits. and licensing management.

Ele directed a number of engineering evaluations and audits including the Design Control Review, INPO Pilot Response, and INPO Self-Initiated Evaluation on the Vogtle Project.

lie has been active in a number of industry groups and AIF and EPRI sIbcommittees and working groups.

Education:

() University of Alabama B.S., Mechanical Engineering l

HOBERT C. SOMMERFELD, Supervising Construction Engineer, l Construction Team Member j O Mr. Sommerfeld has over 20 years of experience in the nuclear power industry. Hi's responsibilities have included design assistance; material selection and performance; corrosion control: preparation of technical specifications: review and approval of vendor and contractor special process procedures and  ;

(.) quality programs; preparation of project QA manuals; interface with regulatory agencies: participation in code writing body 8.6-5

activities; and control of welding operations, forging, heat treatment, and QA/QC activities related to manufacture of pipe, pipe fittings, extruded headers, vessels, and piping subassemblies.

Education:

University of Wisconsin B.S., Mechanical Enoineering CHARLES M. SUMMERS, SF., Administrative Qua'ity Control Engineer, Construction Team Member Mr. Summers has been associated with the nuclear power industry for 7 years in adminirtrative and mechanical quality control.

He has been assigned to various nuclear projects including WPPG Units 1, 2, and 4, Z in me r , Limerick, and Hope Creek. His duties have ranged from inspection of mechanical equipment and piping to QA/QC documents / records control and departmental management of a large constructicn management QC surveillance group which included various discipline activities and construction management task forces.

Education:

Chattanooga State Technical Community College Civil Engineering Technology ZOLLY G. TUCKER, III, Quality Control Engineer, Construction O

Team Member Mr. Tucker has over 5 years of nuclear power plant experience with Bechtel Power Corporation. His responsibilities include supervising, training, and evaluating Level 1 and Level 11 inspectors; and performing all types of in-process and postinstallation in-r etion and testing on piping, instrumentation, welding, and hangers. He is also throughly familiar with the conr.truction turnover, startup, and maintenance phases of a nuclear power facility.

Education:

h Southeastern University 2 years Pre-Engineer Training CARL W. VERNON, Principal Engineer, Design Team Member h

Mr. Vernon's 16-year career has included responsibilities in various disciplines in both construction and operation of nuclear power plants. He has most recently been the Westinghouse project engineer during the reverification and O

8.6-6 l

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f- startup of the Diablo Canyon power plant. He has been involved

(_j in startup and operating procedures, ALARA consultation, plant layout, fluid systems design, NSSS interfaces with balance-of-plant equipment, and analysis of postaccident hydrogen generation. For 5 years, he was assigned to the engineering staff of the U.S. Navy Division of Naval Reactors.

) Education:

Ohio State University B.E.E. Electrical Engineering WILLIAM M. WR IGIIT , Mechanical Project Engineer, Design Discipline Manager Mr. Wright has over 12 years of nuclear power plant experience in mechanical design. He was most recently responsible for managing a group of engineers and pipe designers involved in BOP system design and pipe / pipe support design activities for the Vogtle Electric Generating Plant. He was also involved in several design control evaluations conducted on the Vogtle project which involved technical audit /INPO type reviews of Bechtel, Georgia Power, and Westinghouse organizations.

Mr. Wright was also an engineering group leader for BOP system design activities on Plant Vogtle; a design engineer for developing in-service inspection plans (per ASME XI) on the

% Parley Nuclear Power Plant; and a design engineer on the Barton Nucleat Power Plant where he developed P& ids, system calculations, and hazards analyses for NSSS and safety-related systems and participated in writing the Barton PSAR.

Education:

University of Alabama B.S., Mechanical Engineering M.S., Mechanical Engineering P.E., State of Alabama O

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