ML20134L244

Forwards Response to Request for Info Per 10CFR50.54(f) Re Adequacy & Availability of Design Bases Info

ML20134L244
Person / Time
Site:	Farley
Issue date:	01/31/1997
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9702180380
Download: ML20134L244 (53)

Text

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Dave Morey Southern Nucien Vice President Operating Company Farley Project P0 fiox1295 Birmingham. Alabama 35201 Tel 205 992.5131 SOUTHERN i January 31a 1991 COMPANY y,,,,,,, Sug g,, wig.

Docket Nos.: 50-348 10 CFR .50.4.

50-364 f

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington,DC 20555 Joseph M. Farley Nuclear Plant Response to Request for Information Pursuant to 10 CFR 50.54(f) Regarding Adequacy And Availability of Desien Bases Information '

l Ladies and Gentlemen:

By letter dated October 9,1996, the Nuclear Regulatory Commission (NRC) requested licen-sees to submit information which will provide the NRC added confidence and assurance that j cach licensee's plant (s) is operated and maintained within the design bases, and that any devia-l tion is reconciled in a timely manner.  !

Southern Nuclear fully agrees with the NRC's position that licensee programs to maintain i configuration control should be suflicient to provide reasonable assurance that the plant's ,

physical and functional characteristics are consistent with, and are maintained in accordance  ;

with, the plant's design bases. Southern Nuclear also supports the NRC's belief that licensees are responsible for knowing the plant's licensing bases, having the appropriate documentation that defines the design bases, and providing formal guidance for assessing plant and/or proce-dure changes required by NRC regulations. Southern Nuclear believes that maintaining the design integrity of Farley Nuclear Plant in a reasonable manner, while integrating the design function with operations, maintenance, and license requirements, is consistent with ensuring efficient, safe plant operation, and providing reasonable assurance of protecting the health and safety of the public.

97021803ao 970131 l

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U.S. Nuclear Regulatory Commission - Page 2 l

l Pursuant to 10 CFR 50.54(f), Southern Nuclear hereby submits the enclosed response to the 'l NRC's request to provide information regarding the adequacy and availability of design bases  !

information for Farley Nuclear Plant, in preparing this response, no special design bases veri- l fication review or reconstitution eff' ort was undertaken; however, as discussed in section F,  !

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" Design Review / Reconstitution Programs," a licensing basis assessment, which meets the in- l L tent of NEI _96-05 " Guidelines for Assessing Programs for Maintaining the Licensing Basis," is ,

l being performed. No significant programmatic breakdowns associated with maintaining the

• l- plant licensing and design bases have been identified from this assessment; Commitments to  !

future activities by Southern Nuclear to verify and validate the design bases of Farley Nuclear ,

Plant are discussed in section F.5. j i

Mr. D. N. Morey states that he is vice president of Southern Nuclear and is authorized to exe-cute this oath on behalf of Southern Nuclear. Having read the contents of the Enclosure and, in reliance on the processes described in the Enclosure utilized to develop and validate that Enclosure, Mr. Morey does hereby affirm that the contents of the Enclosure are true and cor- l rect to the best of his inforruation, knowledge, and belief.  !

Respectfully submitted, j SOUTHERN NUCLEAR OPERATING COMPANY

)kk'^ l Dave Morey Sworn to amisubscribed before me this 31 day o$udas 1997

%cfu eWA Notary Public My Commission Expires bvM /TT'7 U

LLB/ cit:desibase. doc Enclosure l l

l cc: Mr. L. A. Reyes, Region II Administrator i Mr. J. I. Zimmerman, NRR Project Manager Mr. T. M. Ross, Plant Sr. Resident Inspector {

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i ENCLOSURE l

FARLEY NUCLEAR PLANT ,

Response to Request for Information Pursuant to ,

10 CFR 50.54(f) Regarding Adequacy and Availability of

Design Bases Information l

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EXECUTIVE

SUMMARY

1 INTRODUCTION 3 A. ENGINEERING DESIGN AND CONFIGURATION CONTROL PROCESSES 5 A.1 Design Change Process 6 A.l.1 Design Change Initiation 6 A.I.2 Design Change Package Preparation 6 A.I.3 Design Change implementation Package Preparation 8 A.I.4 Design Implementation 9 A.I.5 Documentation Updates 10 A.I.6 Temporary Modifications 10 A.I.7 Minor Modifications 11 A.I.8 Modifications in Progress (Work Controls) 11 A.210 CFR 50.59 Implementation Process 11 A.3 10 CFR 50.71(e) Implementation Process 14 A.410 CFR 50, Appendix B Implementation Process 15 A.5 Plant Procedure Control Process 15 A.6 Commitment Tracking 16 A.7 Procurement Process 16 A.8 Document Control Process 17 A.9 Work Control Process 17 A.10 Reload Core Design and Licensing 18 B. TRANSLATION OF DESIGN BASIS REQUIREMENTS INTO PROCEDURES 19 B.1 Initial Design and Licensing 19 B.2 Current Procedure Control Process 20 1 B.2.1 Request for Procedure Development or Resision 20 B.2.2 Procedure Review and Approval 21 B.2.3 Procedure Validation 21 B.2.4 Procedure Approval and Implementation 22 l B.2.5 Maintenance of Procedures 22 l B.2.5.1 Plant Design Control Program 22 B.2.5.2 Operating Experience Program 23

, B.2.5.3 Licensed Operator Requalification/ Simulator Training Program 23 l B.2 5.4 Deficiency Control Process 23 1 B.2.5.5 Licensing Document Changes (e.g., Technical Specification and FSAR Revisions) 23 B.2.5.6 Quality Assurance Program 23 l B.2.5.7 Vendor Documents Review Program 24

! B.2.5.8 Plant Personnel Feedback 24 i B.3 Reviews and Assessments 24 l B.4 Audits and Inspections 30 j B.5 Procedure Use Process 30 B.6 Summary 31 C. CONSISTENCY OF SYSTEM, STRUCTURE, AND COMPONENT CONFIGURATION AND PERFORMANCE WITH DESIGN BASES 32 j C.I Configuration Control Processes 32

, C.2 Significant Inspections 33 C.3 Conduct of Operations 34 Cj Summary 35 i

. i D. PROBLEM IDENTIFICATION AND CORRECTIVE ACTION IMPLEMENTATION ,

PROCESSES 36 l D.1 Deficiency Control Process 36 D.2 Resiew of Potential Defects and Noncompliances 36 D.3 Operating Experience Program 37 D.4 Concerns Program 37 D.5 Safety Audit and Engineering Resiew Program 37 D.6 Occurrence Reporting Program 38 D.7 Root Cause Program _ 38 D.8 Reportability Determinations 38 E. OVERALL EFFECTIVENESS OF CURRENT PROCESSES AND PROGRAMS 39 F. DESIGN REVIEW / RECONSTITUTION PROGRAMS 41 F.1 Proposed Program Assessments 41 F.2 FSAR Accuracy Verification and NEl 96-05 Initiative 42 F.2.1 FSAR Review 44 F.2.2 NEI 96-05 Initiative - Programmatic Sampling 44 F.2.3 NEl 96-05 Initiative - Sarrpling of Potential Changes that May Occur Separate from Pro-grammatic or Procedural Changes 45 F.2.4 NEI 96-05 Initiative - Evaluation 46 F.3 Programs Related to NUMARC 90-12 47 F.4 Safety System Self Assessments 47 F.5 Commitment Tracking Upgrade Program 47 F.6 Specific Commitments 47 F.7 Summary 48 6

Enclosure i i

FARLEY NUCLEAR PLANT Response to Request for Information Pursuant to

10 CFR 50.54(f) Regarding Adequacy and Availability of l Design Bases Information EXECUTIVE

SUMMARY

By letter dated October 9,1996, the Nuclear Regulatory Commission (NRC) issued to licensees a request for information pursuant to Title 10 Code of Federal Regulations (CFR), Part 50.54(f) regarding the adequacy and availability of design bases information.

Licensees were required to submit, within 120 days of receipt of the letter, information that provides the NRC staff added confidence and assurance that plants are operated and maintained within the design bases.

In response to the NRC's request, a working group cognizant of the licensing and the en-gineering design configuration and control processes that support the operation of Farley

Nuclear Plant was assigned responsibility for preparing the submittal. The integrated team approach required the panicipation of corporate, site, and architect / engineer personnel.

Direct leadership for this effort was the responsibility of a corporate oflicer. Supporting levels of management were constantly involved throughout the submittal development process. A multi-level project review was conducted by responsible and knowledgeable individuals involved with maintaining the plant design bases to ensure accuracy and com-pleteness of this response, with perspectives ranging from the Plant Operations Review Committee to the implementation working level.

The focus of this response is principally process oriented with conclusions supported by previously completed procedure and design bases verification activities as discussed in section B, " Translation of Design Bases Requirements Into Procedures," and section C,

" Consistency of Systems, Structures, and Component Configuration and Performance with Design Bases." For the purposes of the NRC request, no special design basis verifi-cation review or reconstitution effort was undertaken, however, as discussed in section F,

" Design Review / Reconstitution Programs," a licensing basis assessment, which meets the intent of NEI 96-05," Guidelines for Assessing Pmgrams for Maintaining the Licensing l Basis,"is being performed. No significant programmatic breakdowns associated with maintaining the plant licensing and design bases have been identified from this rasessment.

Commitments to future activities by Southern Nuclear to verify and validate the design bases of Farley Nuclear Plant are discussed in section F.5.

1 Events, violations, deficiencies and audit findings provide indications of performance and opportunities for improvement. The general results, weaknesses, comments, and/or cor-rective actions resulting from these types of activities provide reasonable assurance that the plant is being maintained and operated safely, and that processes are in place to rec-ognize and resolve problems before they evolve into significant safety issues.

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Based upon the information presented herein, Southern Nuclear concludes that there is reasonable assurance that the current processes and programs at Farley Nuclear Plant are i effective in controlling changes to the design bases and providing reasonable assurance  !

I that the plant's configuration and operation are consistent with its design basis. Southern j Nuclear believes these processes and programs will detect any existing problem or poten-

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tial problem before it becomes significant. Southern Nuclear and the NRC set demanding '

performance standards for people and equipment that may affect safety. The standards are i sufficiently conservative so that a failure to meet them allows the identification and cor- i rection ofproblems before they evolve into circumstances having significant safety conse-  !

quences.  !

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.~ i INTRODUCTION l t

PLANT HISTORY l l

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t Farley Nuclear Plant Units 1 and 2 are Westinghouse three-loop pressurized water reactor  !

designs with each turbine / generator output rated at 860 MW.. Unit I and Unit 2 were issued operating licenses on June 25,1977, and March,31,1981, and commenced com-(

j mercial operation on December 1,1977, and July 30,1981, respectively. The operating licenses were granted with maximum reactor thermal power rated at 2652 MWu . Farley j L Nuclear Plant is located near Ashford in Southeast Alabama.  ;

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j As owner ofFarley Nuclear Plant, Alabama Power Company originally retained respon- .

!' sibility for design and configuration control since the initial construction permit applica-tion. During the design and construction of Farley Nuclear Plant, Alabama Power Com-  !

, pany served as the primary project management organization. Detailed engineering design l

services were provide by three organizations. Bechtel Power Corporation (BPC) was  ;

primarily responsible for design of safety-related systems and structures (other than the '

nuclear steam supply system). Westinghouse provided the nuclear steam supply system.

Southern Company Services (SCS), a wholly owned subsidiary of Southern Company and sister company to Alabama Power Company, was generally responsible for design of non-safety related, power generation systems and components as well as certain safety-related ,

l. systems and components.

Relationships among Alabama Power Company, BPC, Westinghouse, and SCS have been maintained throughout the operating life of Farley Nuclear Plant. Currently, Southern Nuclear Operating Company (" Southern Nuclear"), a sister company to Alabama Power  ;

Company, is responsible for overall configuration control and management of engineering  ;

services provided by SCS, Westinghouse, and BPC. The long-standing relationship be- l tween Southern Nuclear and its primary engineering services providers has helped to en-sure consistency and retrievability of design basis documentation.

NRC REQUEST BACKGROUND By letter dated October 9,1996, the NRC issued to liccasees a request for information pursuant to 10 CFR 50.54(f) regarding the adequacy and availability of design bases in-formation comerns band on the following:

"NRC's findings during inspections and reviews have identified broad pro-grammatic weaknesses that have resulted in design and configuration defi-ciencies at some plants, which could impact the operability of required

! equipment, raise unreviewed safety questions, or indicate discrepancies between the plant's updated final safety analysis report (UFSAR) and the

j. as-built or as-modified plant or plant operating procedures. These inspec-i tions and reviews have also highlighted numerous instances in which timely i

and complete implementation of corrective action for known degraded and E-3

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l nonconforming conditiens and for past violations ofNRC requirements has not been evident, i

1 l The mignitude and scope of the problems that the NRC staff has identified >

L raise :oncerns about the presence of similar design, configuration, and op-l erability problems and the effectiveness of quality assurance programs at l other plants. Of panicular concern is whether licensee programs to main-tain configuration control at plants licensed to operate are sufficient to demonstrate that plant physical and functional characteristics are consistent i with, and are being maintained in accordance with, their design bases."

l l The NRC specifically requested licensees to provide the following information:

a. Description of engineering design and configuration control proc-esses, including those that implement 10 CFR 50.59,

, 10 CFR 50.71(e), and Appendix B to 10 CFR Part 50; 1

b. Rationale for concluding that design bases requirements are translated into operating, maintenance, and testing procedures;

c. Rationale for concluding that system, structure, and component con-

! figuration and performance are consistent with the design bases;

d. Processes for identification of problems and implementation of cor-rective actions, including actions to determine the extent of problems, action to prevent recurrence, and reponing to NRC; and l

e. The overall effectiveness of your current processes and programs m l

l concluding that the configuration of your plant (s) is consistent with {

l the design bases. i l

The NRC further requested that, in responding to items (a) through (e), iicensees indicate whether any design review or reconstitution piagrams have been undertaken, and if not, provide a rationale for not implementing such a program.

l The organization of this response follows the order shown above, as well the Table of I Contents; i.e., letter items (a) through (e) followed by item (f) which addresses the design review program.

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ENCLOSURE I Response to Request for Information Regarding Adequacy and Availability of ,

Drsign Bases Information l I

A. ENGINEERING DESIGN AND CONFIGURATION CONTROL PROCESSES NRC Reauest i

" Description ofengineering design and configuration controlprocesses, including l those that implement 10 CFR 50.59,10 CFR 50. 71(e), and Appendix B to 10 i CFR Part 50. "

Southern Nuclear Response j Engineering Design and Configuration Control Processes The Farley Nuclear Plant engineering design and configuration control prog, am was developed and is maintained to provide reasonable assurance of the compatiallity of the plant's physical and functional characteristics with the design basis and plant documentation. The evolution of system functional requirements, regulatory re-quirements, codes and standards, and plant modifications that has transpired since initial operation requires the control processes to be both integrated successfully and flexible enough to allow changes, while maintaining requirements and protecting the integrity ofdesign and operation. The following primary processes compose the Farley Nuclear Plant design and configuration control program.

• Design Change e 10 CFR 50.59 Implementation e' 10 CFR 50.71(e) Implementation

. 10 CFR 50, Appendix B Implementation e Procedure Development and Revision

• Commitment Tracking

• Procurement

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• Document Control  !

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• Work Control Processes i l

l . Reload Core Design and Licensing l

Each of these processes is discussed in detailin sections A.1 through A.10.

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i ENCLOSURE l Response to Request for Information  !

l Regarding Adequacy and Availability of Desian Bases Information l~

i A.1 Design Change Process  !

l The design change process discussed below is controlled by plant, architect / engineer .

(A/E), and corporate procedures.

A.I.1 Design Change Initiation A change to plant design becomes necessary for reasons such as new or  ;

changing regulatory requirements, equipment problems, equipment improve- ,

ments, and operational improvements. Any employee can identify the need i

! for a change. Plant management reviews a change proposal to determine j l.

scope, purpose, budget, and cost benefit. The review includes consideration of the number and scope of design changes planned for outage and non-L outage work periods. This is considered so that work scope can be scheduled l and managed properly.

I If the change is approved, a design change request is transmitted from the  !

plant to the A/E for design package preparation.

l t A.I.2 Design Change Package Preparation  ;

l The A/E prepares design changes in accordance t ah written procedures that l conform to the requirements of 10 CFR 50, Appendix B, and ANSI N45.2.11.  ;

(Reference section A.4 for fmther discussion of Appendix B implementation.)

Personnel who prepare design changes are appropriately trained and expen-enced. j i

The basic steps in the design change process are:

a. Acquire a full understanding of the scope, need, and purpose of the re-quested design change.

l l b. Identify and assemble the design inputs in accordance with design proce-j dures that must be considered in preparing the design change. In this con-( text, design inputs include the design basis requirements upon which the fi-j nal design is based. Design inputs are identified from the following

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• - Design basis documents l

• Functional system descriptions E-6

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ENCLOSURE

Response to Request for Information  :

Regarding Adequacy and Availability of

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Design Bases Information i I

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• Design criteria e System specifications e Equipment specifications l J

e Codes and standards )

• Regulatory requirements and commitments l

• Calculations l c. Prepare the detailed design change package, including a 10 CFR 50.59 evaluation. (Reference section A.2 for further discussion of the l

10 CFR 50.59 safety evaluation process.)

Many design change packages undergo three separate design stage review meetings (conceptual, detailed, and final) involving corporate, A/E, and site personnel. These reviews help to promote an understanding of the design change so that it meets the purpose of the original change proposal. Site participation typically involves personnel representing operations, engineer-ing support, maintenance, outage and daily planning, security, training, plant modifications, chemistry, and health physics, as required.

d. Identify the documentation affected by 6e design change in accordance with design procedures. The A/E has responsibility for identifying and im-l plementing changes to documentation maintained by the A/E. Plant per-sonnel are responsible for identifying changes to documentation maintained by the site as discussed in section A.I.3. Applicable documentation is re-l quired to be updated and maintained to reflect the "as--built" ctatus of the plant. The types of documentation that may be revised include: i

• Domestic drawings e Vendor drawings e Vendor instruction manuals e Databases e Indices o Calculations e Design criteria

• Equipment specifications i

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ENCLOSURE Response to Request for Information Regarding Adequacy and Availability of Design Bases Information e System specifications e Functional system descriptions

. Final Safety Analysis Report (FSAR) e Other licensing documents The design change package is reviewed and approved by the A/E, and transmitted to the site for implementation.

A. I .3 Design Change Implementation Package Preparation Upon receipt of the approved design change package on-site, the responsible implementation personnel obtain the appropriate reviews and approvals, such as the plant operations review committee and plant management. During the preparation of the implementation package, the design change package and additional information, if appropriate, are routed to affected departments for review. The Plant Operations Review Committee reviews design changes and their 10 CFR 50.59 safety evaluations. Additional actions in this review proc-ess include:

• Identifying testing requirements, training requirements and functional tests to be performed on the installed modification.

. Writing maintenance work orders to perform the installation, including im-plementation plan if required.

• Identifying revisions to programs, processes and documents required as the result of the design change.

In accordance with approved procedures, each site department is responsible for identifying the necessary revisions to their programs, processes, and pro-cedures as the result of the design change to provide reasonable assurance that plant structures, systems, and components are installed and operated in accordance with established design requirements and plant documentation.

The results of the departmental reviews identify the procedures that must be revised prior to placing the modified structure, system, or component in serv- ,

ice after design change implementation. Plant procedures require the identifi-l cation of necessary procedure revisions concurrent with implementation ap-proval. (Reference section A.5 for a description of the procedure control process.)

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, ENCLOSURE i Response to Request for Information l Regarding Adequacy and Availability of i Desian Bases Information l

l A.I.4 Design Implementation ,

7 Plant modifications are scheduled and implemented according to the imple- i f

mentation package in's tructions and requirements using proceduralized work control processes such as clearances, tagging, work orders, and housekeep- l' l ing. After the modifications are complete, a functional test is performed. Ap-L propriate procedures, identified in the affected department review, are pre-pared or revised prior to placing the equipment / system into service in accor-dance with plant procedures. In addition, training is provided for appropriate ,

plant personnel. j Farley Nuclear Plant also has a dedicated design implementation group, Plant ,

Modification Department, whose responsibilities include: 1

. Ensuring activities related to modifications, including design, procure-ment, installation, testing, and close-out, are effectively coordinated  ;

among responsible groups  ;

e Providing liaison between other departments and Corporate Nuclear Sup-  !

port in matters ofplant design l

( e Providmg engmeenng to perform a review of system design changes for i implementation of design intent  !

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l , Ensuring administrative systems and controls are in place for timely com-pletion of required special tests and design change functional tests  ;

. Ensuring modification requests are reviewed for inclusion 'in the modifica- -

tion program and that approved requests are identified, prioritized,

[ scheduled and tracked j

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The Farley Nuclear Plant design change package closure procedure requires a walkdown and review ofimplementing work orders to assure the modification I is complete; the work is satisfactory, critical drawings are updated (i.e. as-built notices are complete), testing required to place the system back into op-eration is complete, functional testing is complete, and required training is complete or scheduled. Near term procedure changes required to operate the

system are completed prior to placing the system in operation; long term pro-

cedure changes (e.g., preventive maintenance program procedures) are com-pleted as necessary.

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ENCLOSURE Response to Request for Information Regarding Adequacy and Availability of Desian Bases Information A.I.5 Documentation Updates The plant notifies the A/E and the appropriate site organization, as necessary, of the implementation of a design change and requests revision to affected documents in order to ensure that documentation changes are made as re-l quired by procedures. The types of documentation that may be revised in-clude:

. Domestic drawings e Vendor drawings e Vendorinstruction manuals e Databases

• Indices -

e Calculations e Design criteria e Equipment specifications e System specifications e Final Safety Analysis Report (FSAR) e Other licensing documents Updating of operations critical documentation is expedited and given top pri- ,

ority. The remaining documentation is updated in accordance with approved i procedures that define a required time frame within which the update process must be completed. Each organization is aware of the documentation for which they have update responsibility. l A.I.6 Temporary Modifications Temporary changes to the physical design ofinservice plant equipment, sys-  !

tems, components, or structures are governed by plant procedures and are  ;

called minor departures (MDs). MDs are reviewed (including a 10 CFR  !

50.59 review) and approved to ensure the licensing and design bases of the plant are not adversely affected during the period in which the MDs are in place. )

! Documentation for an MD includes a description of the change and its effect

on plant operations. The description should be of sufficient detail that Opera-l E-10

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i i ENCLOSUdE Response to hequest for Information Regarding Adequacy and Availability of f Desian Bases Information i

, tions personnel are provided a level of technical detail consistent with the level  !

! of detail provided for other plant design changes.  !

Plant personnel maintain a log indicating the status of active minor departures. .

Operations personnel ensure the installation and removal of minor departures  !

are authorized. Engineering personnel periodically review the status of active  ;

j_ minor departures to maintain awareness of their effect on plant status.  ;

i . i l' A.I.7 Minor Modifications 1 Minor modifications at Farley Nuclear Plant include non-safety related changes, and changes of very limited scope on safety-related equipment and systems. Such changes are referred to as "On-site Design Changes." Proce- '

dures provide controls, including reviews and approvals which are commensu- I rate with the scope, complexity, and safety significance of the on-site design change. Design documents are revised, as necessary, and 10 CFR 50.59 re-views are required, as described in section A.2 below. .

t A. I .8 Modifications in Progress (Work Controls)

I All modifications in progress are governed by work control processes (e.g.,

work orders, clearance and tagging), independent verification requirements, ,

functional testing, and stringent release requirements to assure the design and  !

licensing bases of the plant is maintained. Prior to equipment or systems be-ing released for service, functional tests are performed to determine that the structures, systems, and components function in accordance with predeter-mined criteria. Functional tests may include: surveillance tests, ASME Code Section XI requirements, IEEE Standards service and/or performance tests, and inspections in accordance with appropriate procedures.

- A.2 10 CFR 50.59 Implementation Process Southern Nuclear has established procedures consistent with NSAC-125," Guidelines for 10 CFR 50.59 Safety Evaluations," June 1989. For each proposed modification, a 10 CFR 50.59 evaluation of the change is prepared to determine whether a Technical Specifications change or an unreviewed safety question is involved. The evaluation

- includes a determination of whether or not the activity being evaluated requires a change to the FSAR.

The 10 CFR 50.59 evaluation process applies to virtually all engineering, administra-i tive, and operational change processes, and includes a screening process to determine whether a 10 CFR 50.59 safety evaluation is required for a proposed activity. The I

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l Regarding Adequacy and Availability of (

Desian Bases Information screening process involves evaluating the activity against the description of the plant in  ;

the FSAR and the Technical Specifications. If the screening process indicates the pro- ,

posed changri or activity constitutes a change to the plant or procedures as described I i

in the FSAR., or represents a new test or experiment, a 10 CFR 50.59 safety evaluation i is prepared to cem. iine whether an unreviewed safety question is involved. i

,. A proposed activity that e'.ther constitutes a change to the plant as described in the l FSAR or requires a change to the Technical Specifications is documented and for-  :

warded to appropriate personnel to ensure licensing document changes are made or li-  !

cense amendments are obtained. A change that requires the issuance of an amendment  ;

to the Technical Specifications is not implemented until NRC acceptance is obtained.

Additionally, a change that involves an unreviewed safety question is not implemented until resolved with the NRC.

l NSAC-125 guidelines are utilized for 10 CFR 50.59 evaluations prepared by plant, 4 corporate, A/E, and vendor personnel. Activities for which 10 CFR 50.59 evaluations i must be prepared include, but are not necessarily limited to, the following:

• New procedures, and procedure revisions and deletions e Minor Departures

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I e Design changes, revisions to design changes that impact the original safety evalua- l tion, as-built notices, and licensing document changes e Changes to plans / programs such as the Inservice Inspection Plan Requirements stated in 10 CFR 50.59 establish the conditions unde: an operat-ing license holder may make changes, including tests and experimer . )ut prior NRC approval. NSAC-125 provides guidance for developing prop . ..it consis-tently implement 10 CFR 50.59 requirements.

Documented responses to a series of questions explained in NSAC-125 and based on 10 CFR 50.59 are required by procedure. The responses to these questions determine whether or not the activity requires prior NRC review or approval. Southern Nuclear utilizes NSAC-125 guidance, training of personnel performing the evaluations, and in-dependent reviews to assure thesejudgments are consistently applied.

l l The NRC is informed of the evaluation of changes to the plant or procedures as de-I scribed in the FSAR in a periodic report. This allows the NRC to monitor the effec-tiveness of the engineeringjudgment being applied to these evaluations. Though the j l- NRC has not endorsed the use of NSAC-125, it remains as the guidance to be used l

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ENCLOSURE Response to Request for Information Regarding Adequacy and Availability of Design Bases Information when screening changes or evaluating changes to determine whether prior NRC ap-proval is required. NSAC-125 has proven to be a useful tool that results in conserva- l tive decisions relative to nuclear safety.

Each organization preparing 10 CFR 50.59 evaluations implements control through written procedures that include requirements for the participation of management and supervisory-leve: personnel in the 10 CFR 50.59 evaluation process. Department managers assure that appropriate personnel within their departments are qualified to prepare and review safety evaluations to support activities under their responsibility, and that 10 CFR 50.59 evaluations prepared within their department receive adequate reviews.

l The process for performing 10 CFR 50.59 evaluations is extensively controlled by procedures and requires several levels of review for those changes that involve a change to the plant as described in the FSAR. Some of the key procedural require-  !

raents that assure adequate performance of the evaluations include:  ;

a. The 10 CFR 50.59 evaluation preparer is formally trained on the concepts  !

and terminology intrinsic to tne 10 CFR 50.59 process.  !

b. The safety evaluation reviewer meets the same qualifications that are appli- l cable to the safety evaluation preparer. The reviewer concurs with and en-  !

dorses the safety evaluations. l

c. A plant operations review committee, composed of managers (or alternates) i representing the functional areas of the plant (i.e., a multi-discipline repre-sentation), reviews safety evaluations (indicating a change to the FSAR) pre- I pared according to 10 CFR 50.59 guidelines and concurs with the determi- I nation that an unreviewed safety question is not involved.

d. An off-site nuclear operations review board, composed of oflicers, managers  !

and/or specialists in design, operation, safety analysis or related activities with extensive nuclear plant expertise, reviews safety evaluations prepared according to 10 CFR 50.59 guidelines. Tne purpose of the review is to con-firm the change does not constitute an unreviewed safety question.

l l e. Periodic audits of organizations involved in the 10 CFR 50.59 evaluation l process are performed to verify compliance with established requirements l and determine the effectiveness of the evaluations. Audit results are docu-mented and are reviewed by management having responsibility in the specific areas audited. Follow-up action, including re-audit of deficient areas, is taken as necessary.

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ENCLOSURE l[ Response to Request for Information Regarding Adequacy and Availability of f Desian Bases Information A.3 ' 10 CFR 50.71(e) Implementation Process The FSAR, which was originally submitted to the NRC as part of the operating li-cense application, is maintained and updated in accordance with the requirements of 10 CFR 50.71, and plant, corporate, and A/E procedural guidance. This guid- i ance ensures the periodic revisions, as required in 10 CFR 50.71(e)(3)(i), include the effects of all:

• Evaluations performed in accordance with 10 CFR 50.59 to reach conclusions that changes did not involve an unreviewed safety question, and e Analyses ofnew safety issues.

f As discussed in Section F of this Enclosure, Southern Nuclear is currently conduct-ing an accuracy verification of the Farley Nuclear Plant FSAR.

l The FSAR as submitted to the NRC in accordance with 10 CFR 50.71(e) is avail-able to plant, technical support staff and design personnel for general use and ap-plication. The FSAR is also electronically retrievable for reference purposes.

In addition to the FSAR, other licensing documents such as those listed below are maintained and updated in accordance with plant, corporate, and A/E procedural guidance and submitted to the NRC in fulfillment of a requirement for holding an l operating license:

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• Technical Specifications l'

• Emergency Plan e Security Plan e Off-site Dose Calculation Manual e Process Control Program

. Security Training and Qualification Plan

• Environmental Protection Plan Changes to licensing documents.can result from design changes, as-found condi-l tion evaluations, procedure revisions, and NRC requirements. When a change is proposed, the organization responsible for evaluating the FSAR change determines whether the change will necessitate a revision to required documents, as appropri-ate.

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. 1 ENCLOSURE Response to Request for Information Regarding Adequacy and Availability of Design Bases Information A.4 10 CFR 50, Appendix B Implementation Process

' Organizations involved in the design and configuration control processes described above have developed and implemented quality assurance programs that meet the requirements of 10 CFR 50, Appendix B. The elements of such programs are

! documented in a quality assurance policy manual or document that specifies the l program requirements necessary for 10 CFR 50,' Appendix B compliance, and in procedures that translate these program requirements into day-to-day working guidance.

As required by 10 CFR 50, Appendix B, these implementing procedures provide for the indoctrination and training of personnel performing activities affecting l quality as necessary to assure that suitable proficiency is , -hieved and maintained.

j Procedures among participating design organizations detail the review, approval, l

release, distribution and revision of documents involving design interfaces.

Implementing procedures provide for verifying or checking the adequacy of de-sign, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of suitable testing pro- ,

grams. Design changes are subject to design control measures commensurate with those applied to the original design and are generally approved by the organization that performed the original design, unless otherwise designated.

l A comprehensive system of planned and periodic audits ensures the continuing ef-l fectiveness and adequacy of the design and configuration control processes. These 1 l audits review the programs' continuing compliance with 10 CFR 50, Appendix B, l as well es effective implementation of the program elements.

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! Other 10 CFR 50, Appendix B requirements, such as corrective action measures, control of procurement, etc., are discussed elsewhere in this document. In addi-tion to 10 CFR 50, Appendix B, Southern Nuclear's design and configuration control processes meet the requirements of ANSI N45.2," Quality Assurance Pro-gram Requirements for Nuclear Power Plants."

- A.5 ~ Plant Procedure Control Process The procedure control process provides the method by which appropriate reviews l l

are performed, and design and licensing bases information is accurately incorpo-  !

! rated into plant procedures. Procedure control measures are in place to ensure

procedures used for performing plant activities are evaluated for impact on af-fected processes, documents, commitments described in the FSAR, and organiza-

tions, and are approved by appropriate management. The procedure control proc-4 i

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ENCLOSURE Response to Request for Information Regarding Adequacy and Availability of Design Bases Information ess includes the controls and methods used for procedure development, revision, review, validation, and approval. New or revised safety-related procedures re-quire a 10 CFR 50.59 evaluation as part of the development / revision process. De-velopment of new procedures and revision of existing procedures are necessitated by activities such as design changes, vendor manual updates, operating experience, and industry feedback.

A.6 Commitment Tracking Commitment tracking mechanisms are in place to provide a method ofidentifying commitments and tracking them to completion. Sources of commitments include correspondence to and from the NRC. J A.7 Procurement Process The procurement process is another means by which fidelity of the plant to the de-sign basis is preserved. Procurement of safetj-related items and services for Far-  !

ley Nuclear Plant meets the re.quirements of 10 CFR 50, Appendix B, ANSI N45.2, " Quality Assurance Program Requirements for Nuclear Power Plants," and .

ANSI N45.2.13, " Quality Assurance Requirements for Control of Procurement of Items and Services for Nuclear Power Plants." A Material Engineering organiza- )

tion knowledgeable in these requirements reviews proposed purchases which are l subject to this regulation and standard. l l

Requirements for items and services specified by design organizations are incorpo- 1 rated (or referenced) into procurement documents to ensure that critical character- I istics required for design basis considerations are incorporated into the item or i service received. The A/E may be required to develop detailed specification docu-  !

ments to define engineering and technical requirements applicable to the l

item / service. Each item / service purchased is classified with regard to its safety i significance and appropriate quality assurance program requirements are imposed on the supplier. When required, procurement documents require that the item / service be supplied in accordance with a supplier quality assurance program that has been reviewed and accepted by Southern Nuclear. This review and accep-tance verifies that the supplier has adequate procedares and controls to provide reasonable assurance that the requisite critical characteristics are incorporated into items / services produced.

Upon receipt ofitems, or accomplishment of purchased services, reviews and in-l spections are accomplished to ensure that items / services meet the requirements of

! procurement documents. Periodic audits ensure the continuing efTectiveness and l

adequacy of the procurement control process. .

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ENCLOSURE Response to Request for Information Regarding Adequacy and Availability of Desian Bases Information AE Document Control Process n

Control of documentation includes the preparation, issuance, and revision of de-sign basis documents, licensing documents, and plant documents that specify qual- ,

ity requirements or define activities affecting quality. Document control processes j ensure that documents, including changes, are reviewed, approved for release by authorized personnel, and distributed to the responsible individuals / organizations for implementation and use as appropriate. " Documents," as used in this para-  ;

graph, refers to documents which provide guidance and are subject to revision and as such must be maintained to assure that only the most current revision is used in  :

! activities affecting safety. Categories of such documents include domestic design I

drawings, vendor drawings, plant procedures, plant curves, vendor manuals, and plant manuals. Databases that identify the most recent approved revision of documents are maintained. Controls that prevent the inadvertent use of obsolete or superseded documents are provided. Document control responsibilities include ,

identifying, storing, updating, and retrieving appropriate documents throughout the life of the plant.

A.9 Work Control Process l

( Processes are in place to control work which might othenvise change the plant de-sign or operating configuration. The work is appropriately reviewed, tagged (i.e., a clearance), and tracked to assure that the licensing and design bases of the plant are maintained. Some examples of this type of work are maintenance, surveil-lances, tests, inspections, and modifications. These activities are performed to procedural requirements that specify or describe how the activity is to be per-formed, the plant condition required to perform the work, the methods to be em-played, equipment or materials to be used, and the sequence of operations. Work Orders are the official documentation used to document work performed on plant equipment or systems and are utilized to assure the work history is maintained for tracking and trending purposes. Work orders are also a critical component in im-plementation of 10 CFR 50.65 requirements (the Maintenance Rule), and provide a mechanism for tracking and trending plant equipment performance.

L Work orders are prioritized, planned, tracked, scheduled for work and imple-mented by qualified individuals who review the change for overall impact to the - l plant prior to, during, and after completion of the work. Individuals also review l the work for operability or reportability requirements, ensure personnel safety pre-cautions are identified, and most importantly, consider the design and licensing j bases of the plant by following plant procedures, system operating requirements, and the necessary valve and/or breaker line-ups. Granting clearances, and tagging equipment and components with the proper nomenclature to designate its status is I

E-17 I

i ENCLOSURE Response to Request for Information -

l Regarding Adequacy and Availability of

Desian Bases Information ,

l integral to the work control process. Tags may be used to provide special operat- i ing instructions for equipment that may be out ofits normal alignment, out of i service, or n6t within its operating mode, either by design or for some type of troubleshooting or maintenance activity. The tagging process provides a mecha-l nism to assure configuration control of plant equipment during this time.

l A.10 Reload Core Design and Licensing Prior to each cycle and as a normal pan of the reload core design and licensing process, an evaluation is made to determine or confirm fuel-related safety and op-L erating limits. The design and licensing bases for this process are described in the L

FSAR and its references. Southern Nuclear's fuel depanment procedures require  ;

evaluation of plant changes that could affect reload design and licensing. Prior to each cycle, planned plant design change 5 are reviewed for their impact and inclu- l

sion in the next cycle's reload core design and licensing calculations. Plant changes that would affect reload design and licensing are provided to the fuel vendor using I

a pre-established process. This process provides reasonable assurance the plant ,

configuration is consistent with the reload core design and licensing basis and plant operating limits.

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r ENCLOSURE Response to Request for Information Regarding Adequacy and Availability of Design Bases Information B. TRANSLATION OF DESIGN BASIS REQUIREMENTS INTO PROCEDURES NRC Reauest "Rationalefor concluding that design basis requirements are translatedinto op-erating, maintenance and testingprocedures. "

Southern Nuclear Response- Overview The rationale for concluding that design basis requirements are translated into pro-cedures is based upon the following:

1. Plant procedures were developed based on plant design, the plant as-built condition and the FSAR prior to initial operation.

2. The design, safety related procedures and FSAR have been under procedural control since initial operation to assure that changes to the procedures, the de- ;

sign or to the FSAR are consistent. ,

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3. Audits and inspections are utilized as a means for checking the effectiveness of these processes for maintaining consistency, and in some cases result in defi-  ;

ciencies being written and programmatic or proceduralimprovements. (Audits 4 in accordance with the Quality assurance plan provide a means of checking that the procedures are being followed, and periodic special programs such as l functional inspections, responses to Generic Letters, or implementation of ma- l jor changes such as the Improved Technical Specifications confirm that design i basis, FSAR and procedures are being maintained consistent with each other.)  !

1 B.1 Initial Design and Licensing The original design process produced design documents such as drawings, specifica-tions, evaluations, and analyses necessary to support construction. testing, operation and licensing of the plant. The design documents produced by this process were used as the bases for construction acceptance tests, preoperational testing, and startup testing. The design and construction process incleded various inspections, audits and documentation requirements to assure consistency between construction and design.

The FSAR was produced as part of the design and construction process and submit-ted to the NRC with the application for an operating license. The FSAR includes suflicient descriptions of the plant design to document compliance with regulatory E-19

ENCLOSURE Response to Request for Information Regarding Adequacy and Availability of Design Bases Information requirements and guidance, and specific NRC issues identified during the review of the license application. The plant Technical Specifications were developed in con-junction with the NRC review, and identify functional requirements, controlling pa-rameters, and surveillance and testing requirements that are significant to the safe operation of the plant. Operational requirements were established based on the de-sign and licer. sing information as well as other considerations.

In additicn to verifying consistency between the plant's as-built configuration, the FSAR, and plant design, a preoperational test program demonstrated the erTective-ness of the design control processes and procedures used for initial design and con-struction. Reviews conducted in conjunction with the initial plant startup provided assurance that commitments described in the FSAR were appropriately translated into procedures.

B.2 Current Procedure Control Process The significant steps in the site procedure control process are the request for proce-dure development (new procedure), procedure revision, procedure review, proce-dure validation, and procedure approval. Each of these steps is discussed in detail below.

1 The development of new procedures and existing procedure revisions becomes nec- i essary due to activities such as design changes, vendor manual updates, operating i experience, and industry feedback. Procedure control processes are in place to en-sure procedures used for performing plant activities are evaluated for impact on af-fected processes, documents, commitments described in the FS AR, and organiza- ,

tions, and are approved by appropriate management. The procedure control proc-esses include the controls and methods used for procedure development, revision, review, validation, and approval. The procedure control processes provide the method by which appropriate reviews are performed, and design and licensing bases alennation is accurately incorporated into plant procedures. I B.2.1 Request for Procedure Development or Revision Individuals may initiate a request for development of a new procedure, a pro-cedure revision, temporary change, editorial change or deletion. Each de-partment is responsible for its own respective procedures, tracks the status of these procedures, and coordinates procedure distribution for review and ap-provalin the development and revision process. Each department performs reviews of procedures to ensure consistent format, accuracy and complete-ness. A request for procedure development or revision is normally assigned to the department most affected by the procedure. Someone within the de-E-20

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.- i ENCLOSURE j Response to Request for Information >

-_ Regarding Adequacy and Availability of Desian Bases Information partment is responsible for sponsoring the procedure during the development .

and revision process and preparing the necessary documentation. i B.2.2 Procedure Review and Approval l 3 f

The types of reviews that may be applicable to procedure changes include  !

ALARA, ISI/IST, Emergency Planning, Environmental Qualification, Fire {

Protection, Engineering, Management, Quality Control, Reactivity, Quality Assurance, or Security reviews. These reviews, if required, are performed via l cross-disciplinary reviews by one or more plant organizations. These reviews i l are documented in the procedure request form. The individual responsible for

' the procedure origination or revision determines iflicensing document revi-sions are necessary prior to procedure implementation and if procedure vali-dation is necessary. If the procedure does not require Plant Operations Re-

! view Committee review, the depanment manager is responsible for final ap-proval prior to implementation. Supervisory personnel may approve non-safety related procedures. The Plant Operations Review Committee reviews selected procedures and recommends approval to the General Manager. The General Manager has final approval of these procedures prior to implementa-tion.

For on-site services performed by contractors, procurement documents may require that procedures controlling the services be provided by the contractor for safety-related and other' services. Such procedures are reviewed by cog-L nizant Farley Nuclear Plant personnel, including any applicable cross-

, disciplinary reviews. - A 10 CFR 50.59 review is also performed. Approval of l

the contractor's procedure is indicated by signature of the appropriate plant approval authority.

B.2.3 Procedure Validation The procedure change originator, responsible depanment manager, or Plant Operations Review Committee may identify procedure changes requiring vali-dation during the review process discussed in section B.2.2. When practical, validation of the procedure may be accomplished by actual performance of the procedure. Other methods ofvalidation are simulated procedure perform-

, ance, talk-through, walk-through, comparison with an approved similar pro-l cedure, or a combination of these methods.' If major changes are necessary to

! the procedure or the 10 CFR 50,59 evaluation is revised due to occurrences j during this process, another Plant Operations Review Committee review may

l. have to be obtained.

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. c ENCLOSURE Response to Request for Information

, Regarding Adequacy and Availability of l Design Bases Inf_o_tmation l B.2.4 Procedure Approval and Implementation ,

! After the necessary reviews and assigned open items, as discussed in section l B.2.2, are complete, and comments are incorporated, the procedure is ap-proved for implementation by the responsible department manager or appli-cable plant management. Non-safety related procedures may be approved by supervisory personnel. The record copy of procedures is maintained in i Document Control. .

B.2.5 Maintenance ofProcedures An ongoing dynamic process is inherently required in maintaining procedures

in an accurate and useful condition. This process requires that procedural l

controls be in place for procedure changes as the plant design, regulatory or I operational requirements change.

In addition, many of these procedures are used frequently by plant personnel. l As plant personnel use the procedures, problems are identified and resolved I t

through various internal programs, some of which are listed below. Further, a l significant portion of the normal, abnormal and emergency operating proce-dures are frequently used through various simulator training programs. Once

, identified, proceduralissues are addressed in an expeditious manner. Safety j Audit and Engineering Review audits have indicated that problems in this area have occurred and appropriate corrective actions are being implemented.

Farley Nuclear Plant personnel evaluate procedure maintenance processes and have effective controls to ensure that potential procedural impact is assessed and revisions are made based on input from a number of different programs.

The following programs adequately provide input to procedure revisions and l changes

B.2.5.1 Plant Design Control Program l

l l The plant design control program requires an interface review of modifica-l tions by groups which are potentially affected by the modification. This inter-face review requires that procedures potentially affected by the modification  !

be identified and changes and revisions be ready to be implemented upon completion of the modification. Group managers must indicate that revisions to plant procedures have been issued before the modification package can be considered complete.

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ENCLOSURE i Response to Request for Information  ;

Regarding Adequacy and Availability of l Desian Bases Information B.2.5.2 Operating Experience Program i

The operating experience program requires review of NRC bulletins, notices, .

and generic letters; Westinghouse Owners' Group information; NPO'signifi- l cant operating event' reports (SOERs), significant event reports (SERs), and J operation and maintenance reminders (O&MRs), Nuclear Network operating j plant experience reports; controlled vendor technical information, unsolicited i

, vendor technical information, and various internally generated reports such as  !

l- the Occurrence Report. This review includes an evaluation of applicable pro- I cedures and the initiation of any required procedure changes.

B.2.5.3 Licensed Operator Requalification/ Simulator Training Program l As potential deficiencies are identified in the emergency operating procedures, 1

l formal processes are in place to identify and resolve them. This includes pro-cedure revisions, if appropriate.

l' B.2.5.4 Deficiency Control Process i

l Deficiency control processes are in place so that any individual who identifies

( a potential deficiency reports it to the appropriate authority. As potential de-ficiencies are identified, formal processes are in place to resolve them. This includes procedure revisions, if appropriate.

B.2.5.5 Licensing Document Changes (e.g., Technical Specification and FSAR Revi-L sions) l l

Revisions to Technical Specifications and the FSAR require evaluation for impact on procedures and result in the initiation of procedure changes, if ap-propriate.

B.2.5.6 Quality Assurance Program The Quality Assurance Program includes a review of procedures as part of the audit program. The Quality Assurance Program assigns the responsibility to the Safety Audit and Engineering Review (SAER) group to audit the proce-dural process on a periodic basis. The plant operations review committee also 1

, performs overviews that include a review of many plant procedures. Input  !

l- into the procedure revision process may be provided by either of these two  !

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ENCLOSURE  !

Response to Request for Information j L Regarding Adequacy and Availability of i

'Desian Bases Information B.2.5.7 Vendor Documents Review Program ,

The vendor documents review program requires the review of vendor manuals  !

and revisions to vendor manuals. This review includes an evaluation of appli-  :

i cable procedures and the initiation of any required procedure change. l l

B.2.5.8 Plant Personnel Feedback i

Plant personnel, including operators, are trained and directed by procedures to  !

E report to management any procedural deficiencies or concerns which may ,

prevent or impact their implementation. Feedback into the procedure revision process may be initiated through such programs as the operating experience i

, program.

i B.3 Reviews and Assessments  !

1 To assure that the processes are being appropriately followed and that consistency between the design basis information and the procedures has been maintained, vari-  !'

ous inspections have been conducted since initial operation. These assessments have l provided targeted verification and/or upgrades to the procedures, providing reason-able assurance that the design bases information is adequately translated into proce-dures.

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l g Functional System Descriptions (NUMARC 90-12):

l l In 1989, Southern Nuclear performed a safety system self assessment (SSSA) of the service water system. The SSSA report identified that communications between the designer and the plant staff needed to be improved in the areas ofidentification of design equipment functions (specifically support equipment functions), the identifi-l.

cation of design assumptions and analyses requirements (including the clarification of l: design descriptions in the FSAR), and the documentation of plant decisions regard-l ing design recommendations. To improve under.,tanding of the design bases, Func-tional System Descriptions (FSDs) were developed. As the FSDs were completed for each system, the plant programs were reviewed against the functional require-ments and revised where necessary.

t FSDs were developed between 1990 and 1995 for the following systems:

• Residual Heat Removal

• Safety-Related Electrical Distribution E-24 i

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' ENCLOSURE Response to Request for Information j' Regarding Adequacy and Availability of Design Bases Information j l e Auxiliary Feedwater

. . Chemical and Volume Control

, e Service Water i

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• Control Room Ventilation

! e Post-Accident Sampling

. Instrument Air

. Containment Spray e Component Cooling Water e- Emergency Diesel Generators

. Reactor Protection.

i e Containment Isolation The FSDs documented the system-level functional design requirements, the compo-  ;

! nent and interfacing systern design requirements, and structural design features nec- J l essary to ensure that the system could perform its intended functions. The FSDs i serve as references to the design basis documents for these Farley Nuclear Plant systems. The FSDs are maintained as controlled documents and are revised as nec-essary to ecsure they remain current.

Safety System Self Assessments:

l L Additional SSSAs have been pt.rformed for a total of thirteen Farley Nuclear Plant systems. These assessments were performed by independent teams ofinoividuals

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with design engineering and operations experience who performed an in-depth re-view of system design and operation. Included in the scope of these assessments L was verification of compliance with the design basis described in the system's FSD.

l The process for performance of SSSAs for Farley Nuclear Plant systems was devel-l oped using NRC guidance for SSFIs, in addition to other inputs. The assessments l* included the following systems:

. . Control Room Ventilation e Service Water

! e Residual Heat Removal

1 e Electrical Distribution System and Emergency Diesel Generators

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ENCLOStr ,

Response i. Jequest for Information Regarding Adequacy and Availability of Design Bases Information ,

e Reactor Protection e Safety-Related Electrical Distribution l e Auxiliary Feedwater

• Instmment Air e Post-Accident Sampling

• Component Cooling Water

• Containment Isolation e Containment Spray i 1

• Chemical and Volume Control Findings from these assessments were reponed as strengths or weaknesses. Weak-l nesses resulted in action items which prompted appropriate corrective actions to re-  ;

solve. The weaknesses were documented and tracked to completion of corrective l measures. These weaknesses resulted in submittal of Licensee Event Reports, changes to design basis calculations, design changes and changes to operating, maintenance or testing procedures, as appropriate.

I In Inspection Repon 50-348/96-09 and 50-364/96-09, dated November 8,1996,

! the NRC reviewed the results of certain SSSAs and corrective actions for identified weaknesses, and concluded that the SSSAs were an effective program to identify l and correct deficiencies in design and operation of the systems assessed and that the i corrective actions were appropriate to resolve the identified weaknesses These SSSAs were performed between 1989 and 1995; further assessments are planned.

These reviews provide reasonable assurance that procedures for these systems ade- ,

quately encompass design bases information. l Motor Operated Valve Program:

Farley Nuclear Plant has had in existence since 1987 a specific motor-operated valve l (MOV) program. This component-specific program was initially established in re-sponse to NRC IEB 85-03, and later expanded to include all safety-related and safety-significant MOVs in accordance with the guidance contained in Generic Let-ter (GL) 89-10 and associated supplements. The program is currently under review

, regarding periodic verification testing and trending. This latter review considers the information and guidance contained in GL 96-05. There are currently 354 MOVs included in the scope of the FNP MOV Program.

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l ENCLOSURE Response to Request for Information

~ Regarding Adequacy and Availability of Design Bases Information l' A major element of the program is the verification of the design basis operating re-l quirements for the included valves The verification process identified and con-

~ 1 firmed the pertinent design basis conditions that can exist during requisite valve op-  ;

. eration. Typically included in the design basis evaluation and/or review is:

j e The maximum differential pressure present during valve opening and' closing e The maximum line pressure present during valve operation e The maximum flow rate present during valve operation e ' The required stroke time for valve operation 1 e The worst case available voltage present during valve operation In addition to establishing, as required, the above design basis parameters, a detailed review was conducted to confirm the exact hardware configuration included at each MOV location. The results of the design basis review and hardware confirmation are included in the MOV Setpoint Documents associated with each FNP unit. The setpoint documents are used by FNP personnel as guidance for conducting MOV di- -l agnostic testing and as a reference for hardware-related information applicable to

' the safety-related MOVs at FNP.

1 Environmental Qualification Program:

To meet the requirements of 10 CFR 50.49, a task force reviewed plant systems and developed a master list of electrical equipment required to remain functional during l

and following design basis events to ensure the integrity of the reactor coolant pres-  ;

sure boundary, the capability to shut down the reactor and maintain it in a safe shut- 1 down condition, and the capability to mitigate the consequences of accidents that could result in potential off-sia exposures comparable to the guidelines of 10 CFR g 100. Documentation used for the review included plant piping and instrument dia-  :

grams, plant operating and emergency procedures, Technical Specifications, the FSAR and some emergency procedure guidelines. The master list has been reviewed by various organizations familiar with plant operation and by operations department personnel.

An environmental qualification (EQ) program was established which ensured that a complete review of master list components was performed. The review included l vendor equipment qualification reports, test data supporting vendor EQ test reports  ;

(when available), maintenance procedures, procurement records, differences be- ,

tween tested and installed configurations, and the qualified life. Such reviews are l documented (on a component basis)in environmental qualification packages. These l packages are controlled design documents and are in the document control system.

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l ENCLOSURE Response to Request for Information  !

Regarding Adequacy and Availability of - , l D3 sign Bases Information As a result of these reviews, major EQ program enhancements and upgrades were implemented..  !

The EQ program has been incorporated into appropriate plant operating, make- ,

nance, plant change, and procurement procedures thereby assuring that EQ consid-erations are factored into activities such as maintenance, procurement, warehousmg, document control, and design changes. {

Fuse Program:

' < , A fuse program is being implemented to provide assurance that fuses are used in a controlled manner, and that procedures are in place to ensure design requirements  ;

applicable to fuses are met. This is an on-going program to capture safety-related  :

fuses in a fuse manual. The intent of this manual is to resolve any conflicts in fuse .

sizes and/or types shown on other design drawings, and supersedes such other drawings. This manual identifies the requirements for replacement fuses and identi-  ;

fies acceptable replacement fuses to ensure that the correct fuses are installed when fuse replacement is required. Procedures have been developed to ensure that the fuse manual is updated and revised as necessary.  ;

Power Uprate Project:  ;

i Southern Nuclear is currently preparing a license amendment submittal package for Farley Nuclear Plant Units 1 and 2 requesting an increase in rated thermal power i from 2652 MWa to 2775 MWa for each unit. The Farley power uprate project sup-  ;

porting this submittal has been structured consistent with the methodology estab-lished in Westirg, house WCAP-10263, "A Review Plan for Uprating the Licensed Power of a Preuurized Water Reactor Power Plant," issued in 1983.

The methodology in WCAP-10263 established the ground rules and certain criteria l for power uprate projects, including the broad categories that must be addressed, ,

such as NSSS performance parameters, design transients, systems, components, ac- l cidents and nuclear fuel, as well as the interfaces between the NSSS and the Balance of Plant fluid systems. As a result of the Farley power uprate project, many of the l previous design analyses or evaluations have been reviewed. Specific examples of areas which have been reviewed in detail to address these broad categories include, but are not limited to, large break loss of coolant accident, small break loss of cool-ant accident, fuel / core design, steam generator impact, auxiliary feed water, residual heat removal, component cooling water, service water, containment pres-sure/ temperature response, off-site doses, and others. Inherent in this methodology i

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ENCLOSURE Response to Request for Infonnation Regarding Adequacy and Availability of Design Bases Information are key points that promote correctness, consistency, and licenseability. The key points include the use of well defined analysis input assumptions / parameter values, use of currently approved analytical techniques (e.g., methodologies and computer codes), and use of currently applicable licensing criteria and standards.

In summary, the reviews, analyses, and evaluations performed in support of the power uprate project provide additional confidence that the appropriate design in-formation has been used in the analyses and evaluations forming the design basis of Farley Nuclear Plant.

Improved Technical Specifications (ITS) Conversion Program:

Southern Nuclear is currently in the process of converting the current Technical Specifications for Farley Nuclear Plant to the Improved Technical Specifications (ITS) based on NUREG-1431 for Westinghouse plants. During this process, ex-tensive design and licensing bases reviews are being performed to confirm the appli-cability of the ITS technical specification and bases to Farley Nuclear Plant and to document the basis for that conclusion. The engineering / safety analysis bases for numerical values contained in the Farley Nuclear Plant ITS (operational limits, in-stmment setpoints, required tank volumes, pump flows, etc.) are being identified, verified and documented. A comparison of the surveillanc.: requirements of the ITS with current plant procedures will be performed to identify new requirements and to verify compliance with existing requirements.

Instrument Setpoint Uncertainty Program:

The methods to establish setpoints for safety-related instrumentation have evolved significantly since Farley Nuclear Plant began commercial operation. The initial set-point selection methods were typically based on designer experience and engineering judgment. This approach relied on incorporation of conservative design practices and blanket design margins in the setpoints and the respective analytical and safety limits. However, in 1990 and 1991, licensing activities such as the RTD Bypass Elimination and Vantage-5 Fuel Upgrade required the designers to explicitly ac-count for uncertainties in detailed setpoint calculations. This improved approach provided a documented basis for the uncertainties which could affect the setpoint

such as instmment accuracy, expected environment, calibration variations, and drift.

j This approach also ensured suflicient allowance was provided between the trip set-point and the analytical limit, thereby assuring the associated safety analysis limit (s) would not be exceeded during operational events.

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ENCLOSURE Response to Request for Information Regarding Adequacy and Availability of Design Bases Information Based on these licensing experiences and other recent industry events and informa-tion, Southern Nuclear concluded that Farley should programatically review and en-hance / upgrade the methods and controls for certain safety-related instrumentation setpoints. The resultant program is annotated as the Farley Nuclear Plant Instrument Setpoint Uncertainty Program.

The basic setpoint uncertair.ty program approach for Farley Nuclear Plant is to per-form and maintain uncertainty calculations for the RTS/ESFAS setpoints, NSSS control setpoints associated with the RTDP, and selected setpoints in the Technical Specifications and Emergency Response Procedures. In addition, other uncertainty calculations will be performed and maintained on a case-by-case basis when either requested by the plant staff or determined necessary by the FNP designers. Com-plementary scaling calculations will be typically provided for instrument loops with uncenainty calculations. Administrative controls will ensure that future setpoint changes, instrumentation modifications or replacements, and procedure changes are assessed for instrument setpoint uncertainty and scaling calculation impact. How-ever, uncertainty calculations will only be performed when engineeringjudgment and/or operating experience can not be used to justify the change. Farley's graded-approach to setpoints is generally consistent with IS A recommendations, and the supporting setpoint calculation methodologies are in concert with ISA 67.04 guide-lines.

B.4 Audits and Inspections The Quality Assurance Program includes a review of procedures as part of the audit and surveillance process (which is based on a two year cycle). The Quality Assur-ance Program assigns the responsibility to the Safety Audit and Engineering Review (SAER) group to audit the procedural process on a periodic basis. The Plant Op-erations Review Committee also performs oversight including a review of many plant procedures. Input into the procedure revision process may be provided by ei-ther of these two avenues.

B.5 Procedure Use Process Although there are strict guidelines for procedure origination and adherence, the Farley Nuclear Plant philosophy also maintains a " thinking compliance" with regards to procedures. This philosophy recognizes that procedures cannot be written to cover all situations. Plant personnel rely on the combination of training, procedures, l and peer / supervisor review to successfully accomplish their assigned responsibilities.

As such, Farley Nuclear Plant personnel are expected to consider the intent, and de-sired end result, of procedure usage and provide recommendations for improvement.

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. Response to Request for Information j Regarding Adequacy and Availability of j Desian Bases Information j B.6 Summary l

There is adequate rationale for concluding, with reasonable assurance, that proce-dures are consistent with design and discrepancies among design documents, the FSAR, and procedures are not significant to safety, based on the following: l

• The initial consistency of the plant, procedures and design which was established j during the initial design and licensing process l

• The processes maintaining consistency l

e The various inspections that have verified consistency of design and procedures l l

• Corrective actions, including measures to prevent recurrence, which were under-  !

taken when discrepancies were identified  !

'I e Many years of plant experience and operation l f

' Processes are in place to develop and revise procedures to provide reasonable assur- 1 ance of consistency with the design bases. Processes for identification, evaluation  !

and correction of discrepancies among design documents, the FSAR and procedures .

are also in place and being followed. Discrepancies between documentation and de- (

sign or actual plant conditions are evaluated and resolved as they are discovered. l Plant experience, evaluations and audits of these processes indicate that the desired I results are being achieved.

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l C. CONSISTENCY OF SYSTEM, STRUCTURE, AND COMPONENT CON- l FIGURATION AND PERFORMANCE WITH DESIGN BASES NRC Renuest "Rationalefor conchtding that system, structure, and component configuration arul l performance are consistent with the design bases. "

l Southern Nuclear Response i

The rationale for concluding plant system, stmeture, and component configuration and performance are consistent with the design bases is based on the following:

1. The plant configuration was consistent with design information and the FSAR prior to initial operation.

2. Plant changes have been under procedural control to assure consistency with  ;

the design bases since initial operation. I

3. Routine surveillances, audits, and special inspections check that the processes for maintaining consistency are effective and provide reasonable assurance that plant configuration is being maintained.

4. The plant is operated and maintained in accordance with procedures that in-corporate design basis requirements (as explained in question B).

5. Functional System Descriptions have been developed and Safety System Self l

Assessments have been performed. '

l The activities described below provide assurance that the plant configuration and l

design bases are consistent.

l C.1 Configuration Control Processes l

l Surveillance testing provides a means to verify the operability of systems and com-ponents and ensure variables are within specified design and licensing bases limits.

Surveillance requirements are obtained from Technical Specifications, ASME Boiler and Pressure Vessel Code,Section XI, ISI/IST requirements, and other criteria as necessary. Programs are in place to assure failed or missed surveillances are appropriately evaluated for reportability and corrective actions are imple-

, mented if necessary to prevent recurrence.

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Response to Request for Information l l Regarding Adequacy and Availability of i' l

Desian Bases Information Proposed plant modifications are screened to determine if a safety evaluation is re- i quired per 10 CFR 50.59. Current plant design control processes and the associ- l ated safety evaluations provide assurance that systems, structures and components i remain in agreement with the design and licensing bases of the plant. Change con-trol processes require the plant's systems, structures and components design bases -

information be accurately reflected into design and operating procedures. (See i section A.1 for a detailed discussion of the design change process.)

C.2 Significant Inspections Electrical Distribution System Functional Inspection:

In 1992, the NRC conducted a special Electrical Distribution System Functional Inspection (EDSFI) at Farley Nuclear Plant. The NRC inspection team assessed I the capacity of the electrical distribution system to perform its intended functions, and the adequacy of the engineering and technical support provided to maintain the operability of the system. This inspection consisted of a selective review of design i

calculations, relevant procedures, representative records, installed equipment, and interviews with engineering and technical support staff.

The inspection report concluded that the electrical distribution system and the en-gineering and technical support were generally adequate. Strengths in the report L included the actions undertaken to address DC ground fault detection and emer-l gency diesel generator problems, knowledgeable engineering personnel, good pre-l ventive maintenance, the computer program for emergency diesel generator tran-sient analysis, availability of thermography and time domain reflectometry equip- ,

l ment for predictive maintenance, and on-going actions to upgrade the se vice wa- )

ter system.  !

l Also included were three violations, one deviation, and other items of concern.

Appropriate corrective measures for the violations, the deviation, and other con-cerns listed in the report were taken and appropriate actions to prevent recurrence l l were implemented.  ;

l Service Water System Operational Performance Inspection:

In 1993, the NRC conducted a Service Water System Operational Performance In-spection (SWSOPI) at Farley Nuclear Plant. The NRC inspection team assessed

, the service water system's operational performance; in particular, the team per-  ;

formed detailed reviews of the service water system's design, maintenance, opera-

+

tion, surveillance, and testing. The inspection report concluded that service water E-33 l

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.. 1 l ENCLOSURE Response to Request for Information 1 Regarding Adequacy and Availability of i Design Bases Information  !

system design (excluding a potential water-hammer vulnerability which has subse- l quently been resolved) and operations were satisfactory, partially due to the ag-  ;

}- . gressive follow-up and corrective action to the 1989 Service Water SSSA (see  ;

~'

section B.3).

Three cited and two non-cited violations were included in the inspection report. l l Appropriate corrective measures for the violations and other concerns listed in the i' I

report were taken, and appropriate actions to prevent recurrence were imple-mented.

Safety System Self Assessments:

As discussed in Section B, SSSAs have been performed to provide an in-depth review of system design and operation on thirteen Farley Nuclear Plant Systems, l- and further assessments are planned.

l C.3 Conduct of Operations l

The normal conduct of operations is also in accordance with a philosophy that pre-

serves the integrity of the plant design and licensing bases. The Farley Nuclear l Plant operations department conduct of operations includes, in part, the following

l

• It is fundamental for operations personnel to approach their work with a con-servative mindset, No external factor can be allowed to compromise Southern Nuclear's commitment to safety.

• All senior reactor operators and plant operators on-shift must be aware of and responsible for the plant status at all times.

• Only licensed operators are permitted to manipulate the controls that directly ,

affect the reactivity or power level of a reactor, except for training purposes. l l

l e Whether or not stated in the license, the NRC-licensed individual is subject to I and shall observe all applicable rules, regulations and orders of the NRC.

• All operations personnel are responsible for assuring that the facility is oper-ated safely and within the requirements of the license, Technical Specifications, rules, regulations, and orders of the NRC, and for the actions of their on-duty employees while on site.

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ENCLOSURE l Response to Request for Information Regarding Adequacy and Availability of

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Desian Bases Information i C.4 Summary ,

Adequate rationale for concluding that plant configuration is consistent with design

and that discrepancies between design documents and the plant configuration are i

. appropriately identified and resolved is based on the followmg.

t e The initial consistency of the plant, procedures and design which was estab-i ,

l lished during the initial design and licensing process j e Processes are in place for maintaining consistency, through design control, configuration management, and plant surveillance .

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• The various inspections that have verified consistency of Jesign and proce-l dures  ;

i .

i e Corrective actions, including measures to prevent recurrence, which were un-  ;

dertaken when discrepancies were identified j Processes are in place to monitor sad maintain the consistency between the plant I configuration and performance and the design bases. When discrepancies are identified, there are processes for evaluating and carrecting those discrepancies.

Discrepancies between configuration and design are evaluated and resolved as they

, are discovered. Plant experience, an 3 evaluations and audits of these processes  :

indicate the desired results are being achieved.  !

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Regarding Adequacy and Availability of Design Bases Information i

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D. PROBLEM IDENTIFICATION AND CORRECTIVE ACTION IMPLEMEN- i TATION PROCESSES l NRC Reauest i

"Processesfor identspcation ofproblems andimplementation ofcorrective actions, including actions to determine the extent ofproblems, action toprevent recurrence and reporting to NRC. " 1 Southern Nuclear ResD0nse The primary processes for identifying and determining reportability of problems, and resolving problems are:

D.1 Deficiency Control Process  ;

The deficiency control processes apply to plant-related work concerns and can be i initiated by any employee having direct knowledge of a potential deficiency. Any I adverse condition related to plant equipment, documentation, or personnel safety is normally documented as a deficiency. The threshold for documenting deficiencies  !

is conservative; e.g., normally, a deficiency is documented for any corrective maintenance' activity.

Deficiencies are reviewed for safety significance and reportability. Any deficiency determined to be significant with respect to plant safety or operation receives ad-ditional review in the form of a root cause analysis and identification of corrective actions. If warranted, based upon the importance or complexity of a problem, the investigation of a deficiency may be augmented by establishing a root cause team.

When a deficiency requires corrective action beyond a simple maintenance request or documentation change, corrective actions are typically assigned to individuals in the appropriate depanment, scheduled for resolution, and tracked to closure.

Trend repons are used to identify recurring deficiency-related problems.

D.2 Review of Potential Defects and Noncompliances j The review of potential defects and noncompliances provides a method ofidentify-i ing any items reportable under the provisions of 10 CFR 21, " Reporting Of De-l fects And Noncompliance." Nuclear Steam Supply System (NSSS) supplier corre-spondence, vendor correspondence, and any documented deficiencies on new equipment prior to installation are reviewed for potential 10 CFR 21 reportability.

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D.3 ~ Operating Experience Program The operating experience program assures that operating information pertinent to l plant safety is reviewed and distributed in a timely manner to appropriate plant  !

personnel. Operating information is forwarded to the training department for in- l corpo. stion into the training program as appropriate. The type ofinformation re-  :

viewed for the operating experience program consists ofindustry information, in-cluding NRC bulletins, generic letters, information and enforcement notices; INPO t significant occurrence reports; and NSSS supplier information letters. This infor- l l

. mation is evaluated to determine whether an event is applicable to the plant, and

the review is documented. Any corrective actions are identified and tracked to

( closure.

D.4 Concerns Program l

l l

The concerns program provides a method by which permanent and contractor employees associated with the plant may report any work-related activity or event

( believed to require the attention of management. The anonymity and confidential-l- ity of the individual submitting the concern are maintained, if requested. Concerns .

l are investigated, evaluated for reportability, and documented. A concern is closed  !

l following management review and approval. A response to the concern is pro-vided to the submitter ifidentification was provided.

As part ofinitial employee training, employees are informed of the concerns pro-gram and acknowledge their awareness of the existence and goals of the program by signature. Employees are informed that the program does not affect an individ-ual's right to cany a concern to governmental and/or regulatory agencies.

D.5 Safety Audit and Engineering Review Program The safety audit and engineering review organization conducts audits of plant ac-tivities in accordance with the FSAR and regulatory requirements. Audit fre-quency may increase based upon significant changes in the quality assurance pro-gram or plant perfomiance. Special audits ofimportant industry issues, areas of management interest, and other activities not specifically required by the FSAR may be conducted.

, The audit staff has technical experience in plant functional areas and, when techni-cal expertise is not available on staff, audit teams are often augmented by technical experts from other organizations. Audit results are reported to the Project Vice President. If the audit identifies noncompliances with requirements, the line or-ganization prepares a corrective action response, including root cause determina-J E-37

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Response to Request for Information l Regarding Adequacy and Availability of- l Desian Bases Information  !

L tion if necessary, broadness reviews, and actions required to prevent recurrence.

The response is submitted to the Vice President for approval. Upon completion of  ;

corrective actions, the audit staff reviews the actions to determine their effective- .

ness.

t D.6 Occurrence Reporting Program  !

l l Occurrence Reports are prepared for any occurrence (including personnel errors or  !

equipment failures) which caused, or had a high probability of causing, an adverse l condition that could affect nuclear safety, compliance with regulations or licensing ,

requirements, plant reliability, personnel safety, or plant security. The Occurrence j

Reponing system may be utilized by any employee to document and/or initiate an l

L investigation into the cause of any "near miss" or unusual observation. Such Re-pons are reviewed by appropriate management for reponability determination and

-l to determine appropriate corrective action measures. i l

D.7 Root Cause Program During the operation of Farley Nuclear Plant, there are occurrences which are caused by equipment malfunction / failure or human error. The primary means of 1 identifying significant occurrences and problems are the Occurrence Reporting .

Program and the Deficiency Control Process. These processes provide for root cause determination and corrective action. However, based on the occurrence l significance level, an occurrence may sometimes require a more formal root cause l analysis process, involving commissioning a Root Cause Team. The Root Cause l Team investigates the occurrence and develops a detailed report identifying the' l root cause of the occurrence. Such reports are reviewed by appropriate manage-l ment for reportability determination and to determine appropriate corrective action

!- measures.

l D8 Reportability Determinations Items identified as discrepancies in any of the programs described above are evaluated for applicable reportability requirements per 10 CFR 50.72 and 10 CFR 73.

l Reporting requirements for periodic and non-periodic reports are documented in plant procedures. The majority of problems requiring reporting to the NRC per 10 CFR 50.72 and 10 CFR 73 are identified in programs such as deficiency control processes as discussed in section D.I. However, items identified by other pro-l

grams are evaluated for reportability. Operations department personnel review conditions that are potentially reponable for timely reponing criteria.

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ENCLOSURE l L Response to Request for Information .

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Design Bases Information <  !

E. OVERALL EFFECTIVENESS OF CURRENT PROCESSES AND PROGRAMS  !

NRC Reauest "The overall effectiveness ofyour current processes andprograms in concluding \

that the configuration ofyourplant(s) is consistent with the design bases. " l

)

Southern Nuclear Resoonse l

Southern Nuclear concludes that the current processes and programs are effective in providing reasonable assurance that Farley Nuclear Plant design and configuration is consistent with the design basis based upon the following:

1. The plant design change process provides reasonable assurance of the compati-bility of the plant physical and functional characteristics with the design basis and plant documentation. Changes are evaluated and reviewed to fully determine the

, impact of each Ange on other systems and documents and ensure adherence to

! established requuements. (Reference Section A.1 for a detailed discussion of the design change prccess.)

2. The initial plant verification process, continued consistency control processes, and inspections verify consistency of plant design and procedures. In addition, previously performed design basis verification effons such as SSFI-type inspec-tions and related reviews, represent a baseline that establishes the consistency between the plant, plant procedures, and the design bases. The NEI 96-05 li-

! censing basis assessment has provided initial results; i.e., no significant pro-

, grammatic breakdowns have been identified, thereby providing assurance that structure, system, and component configuration and performance are consistent with the design bases. (Reference Section F for a detailed discussion of the NEI i 96-05 assessment.) l

3. The process ofidentifying and determining reportability of problems and resolv-ing problems includes the deficiency control process, the review of potential de-fects and noncompliances, the concerns program, and the safety and audit and engineering review program. A process is in place to provide assurance that problems are identified, appropriate root cause analyses are performed and, if necessary, corrective actions are implemented. Licensing documents, such as the FSAR, may contain minor discrepancies which should be addressed in due  ;

course. Nevertheless, such a circumstance does not imply that the plant is not consistent with its design bases. Past deficiencies have not been substantial E-39 l

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ENCLOSURE-l Response to Request for Information .

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l enough to cause a lack of confidence in the overall processes and program effec-tiveness.

i

4. Each Farley Nuclear Plant unit has experienced an excellent operating record of  !

l high capacity factors and good SALP and INPO ratings. This experience, along l with other items described in this section, provides reasonable assurance in the l l maintenance of quality standards in configuration control processes that assure  !

the continued adequacy and availability of the licensing and design bases for Farley Nuclear Plant.

l Based upon information presented in Sections A through D of this enclosure, it is concluded that the control processes provide reasonable assurance that the plant is ,

being operated and maintained within its design bases in a mancer that adequately i protects public health and safety. The validity of the current comburation control I processes is continually confirmed by safe and reliable plant operation, on-going I surveillances and walkdowns, self-initiated examinations, and INPO and NRC as- 1 l sessments of plant performance.

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I F. DESIGN REVIEW / RECONSTITUTION PROGRAMS NRC Reauest 1

j Supplemental request for information on design review / reconstitution programs.

l Southern Nuclear Response As stated in our cover letter, Southern Nuclear fully agrees with the NRC's position i that licensee programs to maintain configuration control should be sufficient to pro-vide reasonable assurance that the plant's physical and functional characteristics are consistent with, and are maintained in accordance with, the plant's design bases.

Southern Nuclear also supports the NRC position that licensees are responsible for knowing the plant's licensing basis, having the appropriate documentation that de-fines the design bases, and providing formal guidance for assessing plant and/or pro-

! cedure changes required by NRC regulations. Southern Nuclear believes that main-l taining the design integrity of Farley Nuclear Plant, in a reasonable manner, while l integrating the design function with operations, maintenance, and license require- ,

ments, is consistent with ensuring safe, eflicient plant operation, and providing rea-sonable assurance of protecting the health and safety of the public. To support these concepts, a constant focus on detecting and resolving existing problems or potential problems before they become significant is maintained.

i To implement these concepts, various programs are underway to help provide rea-sonable assurance that the plant's physical and functional characteristics are consis-l tent with, and are maintained in accordance with, the plant's design bases. Such  ;

l programs include:

F.1 Proposed Program Assessments l 1

f Southern Nuclear plans to implement a program to review certain Farley Nuclear L Plant programs to verify and assess the adequacy and accuracy of the implementa-tion.

As has been discussed herein, Safety System Self Assessments (SSSAs) were per-formed to ensure design assumptions were reflected in site procedures and to en-sure design and plant operations are consistent. These SSSAs were very effective in helping ensure design, installation, and operation consistency and adequacy. As  !

a result of recent questions concerning the fire protection program and the piping I support program, it is recognized that a similar evaluation and assessment of vari- l ous FNP programs would be helpful. Therefore a program assessment will be un-1 l

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l ENCLOSURE I Response to Request for Information Regarding Adequacy and Availability of Design Bases Information l

dertaken to increase the level of confidence in selected programs. Southern Nu-clear is in the initial planning stages of this new program evaluation, and will have i a program scope and preliminary implementation schedule developed by April 30, j 1997.

F.2 FSAR Accuracy Verification and NEI 96-05 Initiative In response to NRC Information Notice 96-17, " Reactor Operation Inconsistent I with the Updated Final Safety Analysis Report," dated March 18,1996, and NEI l 96-05, " Guidelines for Assessing Programs for Maintaining the Licensing Basis," a similar assessment as outlined in NEI 96-05 is being completed. An FSAR verifi-cation effort is included in this activity. However, in contrast to NEI 96-05, the j scope of review has been broadened to include the entire FSAR.  ;

Southern Nuclear has conducted an assessment of the programs in place to reaf-firm that Farley Nuclear Plant is being maintained in conformance with its licensing basis. The program met the intent of the program described in NEI 96-05 which  ;

was also included in attachment 3 of the NRC's letter. The program was accom-plished by i

e Reviewing the FSAR

• Sampling programs in place for processing changes to procedures and the plant that may impact the FSAR, and e Reviewing changes that may not be governed by licensee programs.

The programmatic reviews per NEI 96-05 provide assurance that programs are in place and are functioning to identify and correct design and operability problems.

However, the programmatic reviews didn't include an in-depth vertical slice re-view of actual supplementary design basis documentation and a comparison of as-built and as-operated conditions. This type of review was conducted as part of the significant inspections discussed in the response to item C.

The assessments proposed by NEI 96-05 have been completed and final reports are being processed. Southern Nuclear has conducted a preliminaiy evaluation of the results. Specific findings resulting from the audits and reviews are being l evaluated and proposed actions resulting from the lessons learned are being devel-oped. Each of the three areas audited or reviewed are discussed below.

Review guidelines for verifying the information contained in the FSAR were de-veloped, and a division of responsibility was established for reviewing the individ-E-42 l

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o ENCLOSURE Response to Request for Information l

Regarding Adequacy and Availability of l Design Bases Information ual FSAR sections. Personnel determined to be the most cognizant of the infor-mation contained in specific FSAR sections were assigned those sections to review for accuracy. Responsible individuals were directed to focus their reviews on  ;

identifying any discrepancies between the FSAR and the existing plant design, as-l built condition, and operating practices. '

The thoroughness of the review was consistent with the level of detail needed to provide reasonable assurance that the fundamental plant design, as-built condition,  ;

and operating practices are properly described in the FSAR. For example, the re- l view did not require verification of every parameter, such as an instrument set-point, to determine whether it is supported by a calculation. However, if a particu-lar parameter was determined to be suspect based upon the reviewer's familiarity and knowledge of the system, structure, or component, the scope of the investiga-tion was appropriately expanded to seek resolution.

FSAR discrepancies will be corrected in accordance with 10 CFR 50.59 to assure no unreviewed safety question exists and no Technical Specifications changes are required. Currently, no safety significant deficiencies of an operability or report-ability nature have been identified. Accordingly, safety evaluations will be pre-I pared as required by plant procedures, to support the identified changes needed to achieve consistency between the FS AR description of the piant, and the as-built and as-operated condition of the plant. The revisions will be incorporated into the FSAR as part of the normal update process required by 10 CFR 50.71(e).

As outlined in NEl 96-05, a programmatic sampling of activities that involved changes to the Security Plan, Emergency Plan, plant procedures, Technical Speci-fications, and FSAR was performed to ensure that the appropriate licensing bases documents accurately reflect the changes identified in the FSAR verification proc-ess. Based upon the preliminary results of this review, no programmatic deficien-cies that would be considered significant to jeopardize the accurate incorporation ofinformation into the licensing bases have been identified.

In accordance with NEl 96-05, additional sampling of potential changes that may occur separate from programmatic or proceduie changes, such as work around lists, standing orders, old tag-outs, or old temporary modifications, was performed to evaluate the impact on appropriate licensing bases documents. Based upon the preliminary results of this review, no programmatic deficiencies that would be l considered significant to jeopardize the accurate incorporation ofinformation into i the licensing bases have been identified.

The preliminary results of the NEl 96-05 Initiative assessments are outlined below.

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! Desian Bases Information l 1  ;

l F.2.1 FSAR Review i l

l Prior to the decision to perform the assessments of NEI 96-05, SNC had al-l ready committed to an FSAR review program. The program was broader in  ;

scope than the program recommended by NEI 96-05 in that it involved the review of the entire FSAR rather than a selected sample. The review was l I

conducted by assigning each of approximately 600 sub-sections of the FSAR to a cognizant reviewer (in some cases as many as 5 independent reviews .

were assigned). Reviewers were chosen from plant operating staff, SNC, '

SCS, BPC, and Westinghouse. They were assigned sections for review that L were relevant to their positions, experience, responsibilities or projects. They )

were requested to review the FSAR section, identify discrepancies, and initi- l ate appropriate documentation to document their review and any problems l found. SNC is presently evaluating the results of this review. As noted pre-viously a preliminary review indicated that no operability or reportability is- ]

sues have been identified. Once this review is completed, a root cause l evaluation will be performed to determine the cause for identified discrepan-l cies and to provide recommendations for improving performance. j I

F.2.2 NEI 96-05 Initiative - Programmatic Sampling

This audit, performed by Safety Audit and Engineering Review, examined

! changes to the plant, procedures, quality assurance program, Emergency Plan

and Security Plan. It included FSAR changes and changes that were made that required NRC approval. Since these types of changes are screened to determine the degree of applicability of 10 CFR 50.59, the audit specifically included examples of changes that involved a screening that determined that the change did not involve an FSAR or License change, as well as those that

! did require an FSAR change. Approximately thirty-one items were audited.

The primary questions addressed were:

1

1. Should the change have appeared in a licensing basis document?

2. Does the licensing basis document accurately reflect the plant condition?

3. If the answer to 2. is no, then detennine:

i a) the programmatic step intended to ensure accurate incorporation of )

the change, or f

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b) the missing programmatic step needed to ensure accurate incorpo- i ration of the change. I The audit resulted in one comment that identified that no program exists to prompt evaluation of out-of-service non-safety related equipment with re-spect to possible implications for conformance to the plant licensing basis.  !

Consideration is presently being given to establishing guidelines and some l mechanism for evaluation of maintenance or operational practices that could  ;

result in departure from the licensing basis. This comment represents an area of potential weakness, but does not indicate a failure to meet regulatory re-

! quirements.

F.2.3 NEI 96-05 Initiative - Sampling of Potential Changes that May Occur Sepa-rate from Programmatic or Procedural Changes l

The purpose of this sampling and evaluation was to assess the potential for changes that may occur through avenues that bypass or circumvent the pro-grams that are in place to control the implementation of changes. The follow-ing items were reviewed:

• Workaround List e Operations Standing Orders l

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• Several procedures with the potential for "not applicable (N/A)" of steps  !

. The workaround list was utilized to scan for equipment operated in man-ual e Temporary alterations that were over one year old (minor departure list '

used)

• Old non-conformances were evaluated using open commitments from Oc-

! currence Repons (or Incident Reports) and corrective maintenance work requests that are over one year old I . Equipment / systems out of service or tagged-out for over one year A sampling of the items above was selected and reviewed in detail. Research was then conducted to determine whether an impact on the design bases ex-isted, and if so, whether the current design bases /FSAR had been updated to reflect the current condition or practice. Some of the more significant obser-vations of the sampling were:

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1. Two cases were noted where workarounds resulted in operation of l plant systems not consistent with the FSAR. In one case operations

, personnel were providing local manual makeup to the component ,

! cooling water surge tank as opposed to the FSAR description of re-mote manual makeup. The other case was operation of a pressure l control valve in the steam generator blowdown system in manual in-

[ stead of auto as described in the FSAR.

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2. The review of 16 recently completed procedures revealed minor prob-l lems with adherence to the procedure guidelines for marking and justi-fying steps "N/A."

3. A non-safety related systems was found to be tagged out for an ex-tended period of time. Planned maintenance on this systems was not performed. The system, Auxiliary Boiler, was removed from service and is no longer used at FNP. The FSAR had not been revised to re-l flect this condition.

These observations are currently being evaluated to determine the appropriate l corrective actions.

! F.2.4 NEI 96-05 Initiative - Evaluation l

l The " Evaluation Phase" of NEI 96-05 will be performed by Corporate Licens-

! ing personnel. These evaluations will indicate the number of differences that are characterized as being:

. Safety significant

! - Regulatory significant

. Examples where 10 CFR 50.59 was incorrectly applied

. Examples where licensing bases information was always inaccurate In addition the review will:

. Define contributing programmatic weaknesses e Provide rationale for a possible broadening of scope

. Recommend immediate and long term corrective actions l

l E-46 1

i ENCLOSURE 1 i Response to Request for Information

! Regarding Adequacy and Availability of  !

I Desian Bases Information _

The final evaluation of the audits / assessments per NEI 96-05, including the  ;

! FSAR review, and FSAR updates as needed, will be prioritized and imple-  !

l mented in a timely manner.

{

F.3 Programs Related to NUMARC 90-12 Southern Nuclear initiated a program in 1989 to improve configuration con- f trol and design bases information for Farley Nuclear Plant. The program in-  ;

cluded development of Functional System Descriptions and Safety System Self Assessments, as discussed in section B.3. '

F.4 Safety System Self Assessments  !

Farley Nuclear Plant plans to continue its program of SSSAs (see section B.3).

F.5 Commitment Tracking Upgrade Program  !

As discussed earlier in this Enclosure, Southern Nuclear has a program for identify-ing and tracking Farley Nuclear Plant NRC commitments to completion. Southern  :

Nuclear is currently evaluating the NEI " Guidelines for Managing NRC Commit- )

ments" (Revision 2, December 19,1995), and is planning an upgrade to its com-

mitment tracking program. I l

l F.6 Specific Commitments l l

This Enclosure contains much general discussion with the intent of providing in-formation relative to Farley Nuclear Plant's design bases and configuration man-agement control systems to promote a clear understanding of those systems and programs. To ensure a complete understanding of the commitments included l

herein, the following compendium of commitments is provided.

l 1. Two venical slice system assessments and one programmatic assessment will be performed. At least one assessment will be performed in 1997, and the oth-ers performed by the end of 1998. These assessments will include a sampling of the following: conformance to design and licensing requirements, accuracy of controlled documents, e.g. drawings and procedures, equipment walk-downs, operational practices, and maintenance and testing activities. Follow-ing these reviews, the results will be evaluated and a determination will be made by the end of 1998, whether further system or programmatic reviews are

, warranted.

" l

. l E-47 ,

l l l

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+

g ,.

! ENCLOSURE Response to Request for Information i f Regarding Adequacy and Availability of

' Desian Bases Infonnation l

2. Southern Nuclear is currently conducting an accuracy verification of the Farley Nuclear Plant FSAR as described herein. Corrective actions will be identified .

and resolved as previously described. Further reviews will be performed if de- (

termined necessary based on the verification results. l L

3. Southern Nuclear is currently evaluating the NEI " Guidelines for Managing i NRC Commitments" (Revision 2, December 19,1995), and is planning an up-grade to its commitment tracking program. ,

t As with other businesses, Southern Nuclear is constantly evaluating the effective-  ;

ness and continuing need for many of the programs and systems as described herein. In this regard, this Enclosure represents the current state of such programs and systems, with additional historical perspective provided. While compliance with regulatory requirements will be maintained, no information herein should be constmed as a commitment to maintain such programs and systems in place.

F.7 Summary There is reasonable assurance that Farley Nuclear Plant will continue to be oper-ated and maintained in a configuration that is within the limitations demonstrated I

by design, analyses, and industry standards, and will be consistent with safe opera-tion and license requirements. Modification, procurement, operation, and mainte-nance activities will continue to be conducted in accordance with applicable design and licensing documentation, and such information will be maintained correct, ac-cessible, and current. The configuration control processes are designed to assure this policy is being properly implemented.

l 1

L E-48

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