ML17309A872

Provides Info from Util in NRC Ltr Dtd 961009 Re Adequacy & Availability of Design Basis Info

ML17309A872
Person / Time
Site:	Saint Lucie
Issue date:	02/12/1997
From:	Plunkett T FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-97-28, NUDOCS 9702180197
Download: ML17309A872 (150)

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CATEGORY j.

REGULAT INFORMATION DISTRIBUTIONSTEM (RIDE) l ACCESSION NBR'702180197 DOC.DATE: 97/02/12 NOTARIZED: YES DOCKET 0*

FACIL:50-335 St. Lucie Plant, Unit 1, Florida Power a Light Co. 05000335 50-,389 St. Lucie Plant, Unit 2, Florida Power s Light Co. 05000389.

AUTH. NAME AUTHOR AFFILIATION PLUNKETT,T.F. Florida Power 6 Light Co.

RECIP.NAME, RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

. Provides info from util in NRC ltr dtd 961009 re adequacy 6 availability of design basis info.

DISTRIBUTION CODE: A074D COPIES RECEIVED:LTR I ENCL l SIZE: I 0 Responses to 50.54(f) Req. for Design Basis Info t'ITLE:

NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-3 1 1 WIENS,L. 1 1 INTERNAL FILE CENTER 1 1 NRR/DRPM/PGEB 1 1 S,K 3 3 EXTERNAL: NRC PDR 1 1 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-5(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 8 ENCL

Florida Powor Bc Light Company, P.O. Box 14000, Juno Boach, FL 33408-0420 L-97-28 10 CFR 50.4 10 CFR 50.54 (f)

February 12, 1997 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C, 20555 RE't. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 Request for Information Pursuant to 10 CFR 50.54(f) Regarding Adequacy and The purpose of this letter is to provide the information requested from Florida Power & Light Company (FPL) in NRC letter (J. M. Taylor to J. L. Broadhead) dated October 9, 1996. Your letter requested information that would provide the NRC with "added confidence" and reasonable assurance that our plants are being operated and maintained within their design bases and that any deviations identified are reconciled in a timely manner. FPL's response was requested within 120 days of receiving your letter (i.e., by February 13, 1997).

The engineering design, configuration control, and corrective action processes are discussed in the enclosure. These are active processes that are under continuous change and improvement based on operating experience, the corrective action program, and changing regulatory guidance or interpretations.

The enclosed information forms the basis for our conclusion that plant design bases requirements have been and willcontinue to be translated into operating, maintenance, and testing procedures; that system, structure, and component configuration and performance are consistent with the plant design bases; and that FPL has an adequate process for the identification and correction of identified deviations. The overall effectiveness of our current processes and programs provides reasonable assurance that the configuration of St. Lucie Plant Units 1 and 2 is consistent with the plant design bases.

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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Page 2 The Enclosure contains new regulatory commitments which are located in the response to NRC request [fj. In summary, FPL plans to complete the NEI Initiative (96-05) report by April 15, 1997, resolve the 1996 FSAR consistency review findings within a two year period, and complete the plant annunciator summary review in December 1998. Additionally, the recent NRC A/E Inspection findings willbe addressed following issuance of the final inspection report, and the corrective actions will be provided by a separate letter or initiative as appropriate, The enclosed information is provided pursuant to the requirements of Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f). The required oath or affirmation affidavit executed by my Engineering Vice President, is attached.

Please contact us should you have any questions on the enclosed information.

Very truly ours, T. F. Plunkett President Nuclear Division TFP/GRM Enclosure cc: Director, Office of Nuclear Reactor Regulation, USNRC Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Page 3 STATE OF FLORIDA COUNTY OF PALM BEACH R. S. Kundalkar being first duly sworn, deposes and says:

That he is Vice President, Engineering, Nuclear Division, of Florida Power S Light Company, the Licensee herein; That he has executed this document; that he has read the contents of the attached Enclosure, and in reliance on the processes discussed in the Enclosure and independent oversight of its accuracy, hereby affirms that the statements made in this document are true and correct to the best of his knowledge, information and belief; and that he is authorized to execute the document on behalf of said Licensee.

R. S. Kundalkar STATE OF FLORIDA COUNTY OF Sworn to and subscribed before me sids j+ day of Wr /',

by R. S. Kundalkar, who is personally known to me.

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St. Lucio Units I and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page I Response to NRC 10 CFR 50.54(fj Information Request Adequacy and Availability of Design Bases Information

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 2 Table of Contents 1.0 Introduction .................... ~ ~ ~ ~ ~ ~ ~ ~ ~ 3 2.0 Response Preparation ............. ~ ~ ~ ~ 3 3.0 Summaty of Conclusions .......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 4.0 Historical Background . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 4 5.0 L1st of Actloils A

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ 7 NRC Request [a] "Description of engineering design and configuration control processes, including those that implement 10 CFR 50.59, 10 CFR 50.71(e), and Appendix B to 10 CFR Part 50."......................... 8 NRC Request [b] "Rationale for concluding that design bases requirements are translated into operating, maintenance, and testing procedures." ............ ~ ~ 29 NRC Request [c] "Rationale for concluding that system, structure, and component configuration and performance are consistent with the design bases." .. 43 NRC Request [d] "Processes for identification of problems and implementation of corrective actions, including actions to determine the extent of problems, action to prevent recurrence and reporting to NRC."..................... 65 NRC Request [e] "The overall effectiveness of your current processes and programs in concluding that the configuration of your plant(s) is consistent with the d esign bases.............................................

1 tt 76 NRC Request [f] "Supplemental request for information on design review/basis programs."

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 85 Appendix A; Partial List of Acronyms ....................................... 88 Appendix B ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 89 Table B 4.7.1- QA Audit Findings....,............................... 89 Table C 3.12- QA Audits and Vertical Slice Reviews ......................... 93

0 St. Lucio Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 3 1.0 Introduction By letter dated October 9, 1996, the Nuclear Regulatory Commission (NRC) requested information to provide added confidence and assurance that Florida Power & Light Company's (FPL) nuclear units, St. Lucie Units 1 and 2 and Turkey Point Units 3 and 4, are operated and maintained within their design bases and that any deviations are reconciled in a timely manner.

The purpose of this enclosure is to provide the specific information requested by the NRC for St.

Lucie Units 1 and 2. Although the engineering design and configuration control processes are similar for the St. Lucie and Turkey Point units, their licensing and design bases are different and certain activities associated with maintaining the design bases are plant specific. Therefore, the requested information for the Turkey Point units is provided in separate correspondence.

2.0 Response Preparation The information provided in the following enclosure is organized consistent with the five (5)

Requests [a] through [e] of the NRC's October 9th letter, with St. Lucie future actions provided in Request [f]. Additionally, a partial list of acronyms used in this enclosure along with their definitions is provided as Appendix A and the details of selected FPL Quality Assurance (QA) audits as Appendix B. Where the information provided to the Requests [a] through fe] is the same, the information provided in the initial location is later referenced, in context, for the other applicable questions. References made to plant programs, procedures or inspections, are either contained in FPL docketed correspondence or retained onsite.

This enclosure was developed by a multi-disciplined team representing the St. Lucie Engineering, Licensing, Quality Assurance, Configuration Management, Training, and Corrective Action Departments. An interface between St. Lucie and Turkey Point was established to aid in preparing the responses for common areas. Plant and FPL staff management provided guidance and supported the development of the response by participating in team meetings and by reviewing the response outline and drafts. The majority of tasks involved the review of historical documents that describe processes and results of Final Safety Analysis Report (FSAR) reviews and Design Basis Document (DBD) development. Quality Assurance audits of these activities, as well as the processes for controlling configuration and design control, were also reviewed. The process used in the development of this response was similar to preparing an engineering design product, in that, there were response preparers, independent verifiers, and approvals by knowledgeable plant and staff personnel. This process helped to ensure that the necessary information was provided to the Requests, and that the information submitted is complete.

Due to the evolutionary nature of each of the processes discussed in this response, descriptions of processes are a snapshot in time. Each of the processes is subject to controlled change consistent with FPL procedures.

St. Lucio Units 1 and 2 Docket Nos. 50-335 and 50-389 L 97-28 Enclosure Page 4 3.0 Summary of Conclusions FPL understands the importance of the plant design bases and is committed to maintaining the plant consistent with the design bases. FPL has concluded with reasonable assurance that its processes ensure that the as-built plant is an accurate reflection of the plant's design bases and therefore, that plant changes do not erode or compromise the safety margins associated with risk-significant systems. Where performed, in-depth vertical slice reviews of design bases documentation and their use in plant modifications, procedures, and operations, support the overall conclusion that FPL processes are effective in maintaining the St. Lucie design bases.

FPL has reviewed the results of St. Lucie audits, inspections, self-assessments and vertical slice reviews, current engineering design and configuration control processes, plant configuration and performance documentation, and problem identification/corrective action programs. Based on these reviews, FPL concludes with reasonable assurance that the St. Lucie units are operated and maintained within the design bases and that any deviations are reconciled in a timely manner.

Further, it is concluded with reasonable assurance, that the design bases for St. Lucie are accurate, that the design bases and as-built plant configuration and procedures have been maintained consistent with the license through reasonably effective processes, and that the plant corrective action programs adequately identify and correct deficiencies in a timely manner.

FPL has endeavored to keep pace with the latest thinking of the industry and the NRC on design bases and configuration management programs. By doing so, FPL has been able to enhance its programs and processes over time, improving the design bases information and further ensuring plant operation and configuration is consistent with the design bases. The necessary confirmation of this has been provided by the many internal and external assessments that continually occur at St. Lucie. While these assessments have confirmed an overall design bases program adequacy, they have also generated corrective actions to remove identified weaknesses. These types of assessments provide for a continual strengthening of the overall design bases and plant operation.

One recent example of such an assessment is the comprehensive NRC Architect/Engineer (A/E)

Inspection. This inspection concluded that there were no major weaknesses in the St. Lucie design bases information for two major plant systems. Nevertheless, some improvement actions have been identified from the inspection. Similarly, the St. Lucie management initiated self-assessment, conducted in the first half of 1996, found a general adequacy in the design bases area, but identified weaknesses which must be corrected to enable St. Lucie to regain its former excellent performance.

4.0 Historical Background The original St. Lucie Unit 1 FSAR was submitted to the NRC on March 5, 1973. The original St. Lucie Unit 2 FSAR was submitted to the NRC on March 24, 1980. Revisions to the FSARs were submitted throughout the licensing process until the original NRC Safety Evaluation Reports were issued. The design of St. Lucie Unit 2 included enhancements from previous industry events such as the accident at Three Mile Island Unit 2 and the fire at Browns Ferry. These

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 5 enhancements were added to St. Lucie Unit 1 as plant modifications performed in accordance with 10 CFR 50.59, "Changes, Tests and Experiments," and 10 CFR 50.90, "Application for Amendment of License or Construction Permit." Similarly, FPL submitted the Topical Quality Assurance Report (TQAR). Periodic updates to the FSARs and the TQAR are performed in accordance with the Code of Federal Regulations.

During the design and construction phases of St. Lucie, FPL performed project management functions. The A/E (EBASCO) and NSSS supplier (Combustion Engineering) performed the engineering and modification control functions, including original design and any subsequent modifications in accordance with 10 CFR 50.59. Following the implementation of plant upgrades at Turkey Point in response to NUREG 0737, a review of the engineering design and configuration management processes at Turkey Point indicated that the A/E and plant interface was not as robust as during the construction phase. In response to the activities at Turkey Point, St. Lucie implemented the Commitment to Excellence Program (CEP). The CEP did not include a formal design bases reconstitution program, since the vintage of the St. Lucie Plant allowed the capture of most of its design bases information within the original plant documentation.

However, the CEP did lead to improvements to the operation of the St. Lucie units.

Beginning in 1989, FPL began to assume the A/E design functions for both St. Lucie and Turkey Point. St. Lucie design engineers obtained training from EBASCO and worked hand-in-hand with the A/E design engineers in the preparation of design change documents. By late 1990, FPL Engineering was performing design, production engineering, and design modification functions.

The configuration and design control management processes were revised to reflect the in-house process. FPL has concluded that this change was a significant step toward better understanding and controlling the plant's design bases.

In 1991, a configuration management manual was issued which delineated design and configuration control responsibilities as well as the process requirements for plant design, interfaces, reviews and approvals. Engineering and Administrative Quality Instructions were developed to implement design and configuration control process requirements.

In the 1990 time frame, St. Lucie participated in the Combustion Engineering Owners Group project that resulted in the development of the St. Lucie Design Basis Reference System (DBRS).

During the development of the DBRS, it was concluded that a program was needed to develop summary level documents which would consolidate information contained in reference design bases sources (i.e., calculations, analyses, etc.), thereby making it more accessible to respond to plant operational needs.

In the 1991 time frame, in response to the above need, the St. Lucie Design Basis Program was initiated on a number of significant plant systems to develop the summary level documents, referred to as Design Basis Documents (DBD). The St. Lucie design basis program did not involve reconstituting the design bases, however, specific reconstitution efforts have been performed such as the LOCA containment re-analysis and electrical distribution system bases calculations.

St. Lucio Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 6 In addition to the above chronology of efforts, St. Lucie has been involved in various activities which have helped to improve and/or validate the plant's design bases and its design and configuration control processes. These activities include:

In response to NRC Bulletin 79-14, "Seismic Analysis for As-Built Safety-Related Piping Systems," FPL perfor'med walkdowns of designated Seismic Category I safety-related piping systems and reviewed documentation of the as-built configuration.

In 1981, St. Lucie Unit 1 provided analyses to justify an increased thermal power rating.

In response to NRC Generic Letter 87-02, "Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors," (USI A-46), FPL performed an evaluation for the seismic adequacy of selected equipment using FPL engineers and a Seismic Review Team consisting of engineering experts in the area of seismic adequacy.

In 1988, the FPL Quality Assurance Department conducted a vertical slice audit on the St.

Lucie Unit 2 Intake Cooling Water system design.

In 1991, the FPL Quality Assurance Department conducted a vertical slice audit on the St.

Lucie Unit 1 and Unit 2 feedwater and main steam systems.

In February and March 1991, the NRC conducted an electrical distribution system functional inspection.

In 1991, FPL performed a technical assessment of the containment penetration boundaries for St. Lucie Units 1 and 2.

In September and October 1991, the NRC conducted a service water operational performance inspection.

In April 1993, FPL updated the containment Loss of Coolant Accident analysis.

In 1996, FPL performed a Final Safety Analysis Report review for procedural consistency.

From November 1996 through January 1997, the NRC conducted an Architect/Engineer inspection which involved two major systems at St. Lucie; Unit 1 auxiliary feedwater and Unit 2 component cooling water.

In 1997, St. Lucie willbe performing a Unit 1 steam generator replacement. In the course of preparing for this activity, plant walkdowns have been conducted and FSAR, design analysis, and other applicable design bases documents have been reviewed.

The above described chronology and activities, coupled with the improved design and configuration control processes, provide reasonable assurance that the St. Lucie design bases are

St. Lucio Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 7 adequate and that the plant is configured and operated in accordance with its design. In addition, actions are in progress or will be initiated which will further improve the design bases documentation and the processes which control design and configuration of St. Lucie.

5.0 List of Actions St. Lucie has several planned and on-going actions which will provide additional assurance that the units are operated and maintained within the design bases. These actions are discussed further in responses to Request [fJ and are listed below:

5.1 The NEI Initiative (96-05) report willbe issued to NEI by April 15, 1997.

5.2 The 1996 FSAR consistency review findings willbe resolved within a two year period.

5.3 The plant annunciator summary review is currently scheduled for completion in December 1998.

5.4 The recent NRC A/E Inspection findings will be addressed following issuance of the final inspection report, and the corrective actions will be provided by a separate letter or initiative as appropriate.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 8

[a] "Description of engineering design and configuration control processes, including those that implement 10 CFR 50.59, 10 CFR 50.71(e), and Appendix B to 10 CFR Part 50."

The following outline provides the organization of the response to request [a].

1.0 Configuration Management Overview 2.0 Detailed Descriptions of Engineering Design and Configuration Control Processes 2.1 Engineering Design Control Processes 2.2 Configuration Control Processes 3.0 Implementation of 10 CFR 50.59 3.1 Unreviewed Safety Question Determination Process 3.2 10 CFR 50.59 Screening and Evaluation Process 3.3 10 CFR 50.59 Safety Evaluation 3.4 Non-10 CFR 50.59 Activities 4.0 Engineering Design and Configuration Control Process Training 4.1 Engineering Support Personnel (ESP) Training and Qualification 4,2 Shift Technical Advisor (STA) Training Program 4.3 Nuclear Plant Supervisor (NPS) Training Program 4.4 Facility Review Group (FRG) Training Program 5.0 10 CFR 50.71(e) Implementation Process 5.1 FSAR Updates 5,2 Design Basis Document Updates 6.0 10 CFR 50, Appendix B Implementation Process St. Lucie recognizes that maintaining current and accessible design documentation is important to ensure that, (1) the plant physical and functional characteristics are maintained and are consistent with the design bases as required by NRC regulation, (2) systems, structures, and components can perform their intended functions, and (3) the plant is operated in a manner consistent with the design bases. This is accomplished through the St. Lucie Configuration Management (CM) processes, which encompass engineering design and configuration control.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-2S Enclosure Page 9 1.0 Configuration Management Overview Configuration Management (CM) is comprised of processes that are used to control the plant design basis, design, design changes, physical configuration, operations, maintenance, testing, procurement, installation, training and documentation. The CM program goals are to ensure that, throughout the life of the plant the following occur:

1. Design bases are identified, documented and maintained.

2, Approved design documents accurately reflect and implement the design bases.

3. Plant physical structures, systems, components, (SSCs), and process control computer software conform to the approved design requirements.

4. Plant physical and functional characteristics are accurately reflected in plant documents.

5. Plant conditions conform to the requirements included in the design bases, procedures and other approved documents which are used to control the operation of the plant.

6. Changes to the design or physical plant are optimized through an integrated review process with established approval criteria to make necessary changes to the design or plant.

7. Changes to the design are approved, planned, budgeted and scheduled for all phases of the change from design through implementation, required testing and document update.

8. Consistency is maintained between design bases, design documents, hardware, software, and plant procedures.

Activities which affect the CM goals listed above are controlled by engineering design and plant configuration control processes. These processes ensure that the plant design, as-built condition, and operation remain in conformance with the original plant design bases and subsequent licensing commitments and requirements.

2.0 Detailed Descriptions of Engineering Design and Configuration Control Processes 2.1 Engineering Design Control Processes 2.1.1 General Description The processes whereby the design and design bases are maintained and controlled includes all plant departments. The engineering design control processes are controlled under FPL Engineering Quality Instructions (ENG Ql). ENG QIs are designed in a hierarchical fashion with reference to interfacing instructions. The overall governing ENG QI for the engineering design control processes is ENG QI 1.0, "Design Control." This instruction leads the engineer/designer/user to other governing instructions for maintaining and controlling the plant license, design bases, as-built design documentation, and operations, maintenance and testing procedures.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 10 Determination of the appropriate process vehicle to initiate a design change requires the understanding of the scope of the proposed change. The design change process is divided into Plant Change/Modifications (PC/M) and Non-Plant Change/Modifications (Non-PC/M).

The PC/M process is further subdivided into major and minor changes. Major design changes to the plant are controlled under ENG QI 1.1, "Engineering Packages," and are implemented using the complete PC/M Engineering Package (EP) process to perform modifications which constitute a change to the facility (physical and/or documentation), requiring a 10 CFR 50.59 safety evaluation. EPs usually contain detailed implementation and testing information for complex or involved design changes.

Minor design changes to the plant are controlled under ENG QI 1.2, "MinorEngineering Package", and are implemented using the process to perform minor modifications which constitute a change to the facility (physical and/or documentation) but which do not require a 10 CFR 50.59 safety evaluation. Minor Engineering Packages (MEP) include 10 CFR 50.59 screening to document that a 10 CFR 50.59 safety evaluation is not required.

The Non-PC/M process includes Drawing Change Requests (DCR), Item Equivalency Evaluations (IEE), Condition Report dispositions, and Maintenance Specifications. This process also includes Temporary System Alterations (TSA), design document updates, minor repairs, and maintenance activities.

2.1.2 Plant Change/Modification (PC/M) Process For PC/Ms, the engineering design change control process includes:

Design Change Initiation Design Change Package Preparation Design Implementation Documentation Updates 2.1.2.1 Design Change Initiation A change to the plant design may become necessary for a number of reasons, such as new or changing regulatory requirements, equipment reliability problems, obsolete equipment, or operational improvements.

A plant modification is usually proposed by using the Request for Engineering Assistance (REA) process. The REA process, AP 0005745, "Request for Engineering Assistance," can be initiated by any member of the plant staff requiring assistance from Engineering to resolve plant problems, conduct engineering studies, reconcile discrepancies, and update design documentation. This process provides an anticipated scope, purpose, budget, cost and benefit, design/safety significance, and relative priority for the candidate project. It also identifies the sponsoring organization for the proposed modification. REAs are processed using one of two paths: Real-

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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L 97-28 Enclosure Page 11 Time Support Team (RTST) review for easily implemented projects, and Plant Review Board (PRB) management review group for larger or construction-intensive projects. Following the issuance of an REA, Engineering performs a scope, schedule and cost estimate for a proposed project, including the evaluation of design options. Ifvarious options are possible, plant input is solicited so that issues of cost, simplicity, maintainability, usability, etc., may be factored into the evaluation for the preferred option. This initial information is used by the RTST or PRB review groups before they decide to recommend final approval for Engineering to prepare a design package and for the implementor to complete procurement and implementation activities. The review groups are composed of plant management or their designees who review and approve projects, authorizing the use of Engineering resources both for estimating and designing activities.

The review includes consideration of the number of design changes planned for outage and non-outage work period, so that work scope can be scheduled and managed properly. Ifapproved, the REA is forwarded to Engineering for design change package preparation.

2.1.2.2 Design Change Package Preparation The design engineers prepare design changes in accordance with the license, design basis, design standards, and configuration management requirements governed by ENG QI 1.0, "Design Control," and ENG QI 1.1, "Engineering Packages," and ENG QI 1.2, 'Mnor Engineering Package." These procedures are approved under the QA 10 CFR 50 Appendix B program.

Personnel who prepare design changes are appropriately trained and experienced.

The basic steps in the design change process include:

a. The design engineer(s) conducts research and interfaces with the design change sponsor(s) to obtain a full understanding of the design change justification, scope, licensing/design impact, interdiscipline design requirements, affected engineering/operations documentation, etc.

Applicable Operating Experience Feedback (OEF) (lessons learned from other utilities) is factored into designs or design modifications. Engineers stay current with FPL and nuclear industry operating experience related to engineering activities. NRC Bulletins, Information Notices, other regulatory requirements, internal FPL correspondence related to nuclear design experience, experience reports received from manufacturers and other industry sources are circulated appropriately. In the development of a design document, the following computerized information retrieval services are currently available and may be used to obtain data related to design or safety analysis input: OEF program (FPL), Nuclear Plant Reliability Data System (INPO), Technical Library Database (INPO), Nuclear Network (INPO), LER Database (INPO), and the NRC Index (INPO).

b. The design engineer(s) identifies and assembles the design input that must be considered in preparing the design or design change. Design input is defined as the set of criteria, parameters, bases or other information upon which detailed final design is

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 12 based. Design changes are based upon existing design bases documentation. Design inputs are identified from the following sources:

~ FSAR/I'echnical Specifications

~ Regulatory Requirements

~ Licensing Commitments

~ Bases Documents (calculations, diagrams, drawings, design specifications, analyses, studies, vendor documents, DBDs, etc.)

Applicable Codes and Standards Requirements

~ Equipment/Component Specifications

~ Engineering Documents

~ Requests for Engineering Assistance Change Request Notices

~ Condition Reports The design organization performs design integration activities to ensure that all new designs/design changes are consistent with the base plant design, and that consideration is given to related design and design changes in progress. The design integration activities include a review of information sources such as TSA logs, and abandoned equipment logs. To aid the design integration, electronic information sources which are part of the PassPort database (FPL's computerized management information system) are also available. These include the PC/M Index, Calculation Index, Engineering Evaluation Index, and Total Equipment Data Base. This review provides assurance that no other modifications which are in progress can affect the modification package which is being prepared. A limit is imposed by management on the number of modifications that are processed at the same time in order to minimize integration difficulties.

c. The design engineer(s) performs the detailed design change and prepares 10 CFR 50.59 screening and/or safety evaluations.

The PC/M defines the scope of the modification and activities associated with the change. The safety evaluation or screening is a major part of the PC/M and is required to demonstrate that the provisions of 10 CFR 50.59 are met when performing a design modification. The requirements of 10 CFR 50.59 are a subset of the overall safety evaluation objectives. The safety evaluation addresses the overall safety aspects of the modification including description and purpose of the change, analysis of effects on safety, failure modes and effects analysis, effects on Technical Specifications, Unreviewed Safety Question (USQ) determination, and plant restrictions.

The detailed design effort includes walkdowns of affected systems and components to verify their field configuration and their consistency with plant design drawings.

Design change packages undergo interdisciplinary reviews, i.e., conceptual, detailed, and final, involving Engineering and plant personnel. A conceptual review meeting

S St. Lucio Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 13 may be held for larger more complex design changes to solicit suggestions/input prior to commencement of a detailed design. When the proposed design is substantially complete, a draft design change package is issued for Implementation Review (IR) so that input can be solicited from the appropriate departments or groups such as Operations, Maintenance, System and Component Engineers, Health Physics, Configuration Management, etc. The early and final review processes ensure that the design change is understood and meets the design basis, design and operations requirements. They also ensure that the impact of the design change is reflected in the appropriate plant design documents and procedures, and that the final product willbe acceptable upon turnover to the plant. After incorporating interdepartmental comments, the design change package is finalized by Engineering.

d. The design engineer(s) and other affected site organizations identify and assemble the documentation affected by the design or design change for inclusion in the PC/M package. Documentation is updated and maintained to reflect the "as-built" status of the plant. Responsibility for updating each type of documentation is assigned to the appropriate department, functional group, or organization. The PC/M affected design documents are tracked to aid in design integration. The types of documents that may be affected include:

~ FSAR Licensing Documents

~ Design Bases Documents

~ Engineering Drawings

~ Design Specifications

~ Equipment Specifications

~ System Specifications

~ Calculations

~ Analyses

~ Vendor Drawings

~ Vendor Instruction Manuals

~ Databases

~ Indices The final design change package is approved and issued by Engineering management and is processed by the plant for implementation and document control.

2.1.2.3 Design Implementation The design change package is then submitted to Configuration Management. The approved design change package is routed to affected plant departments for review and preparation for implementation. The appropriate department identifies training requirements and any additional functional tests required to be performed following the modification. The implementing departments prepare work documents, i.e., Plant Work Orders (PWO), which provide instructions

St. Lucio Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 14 for implementing the design change. Each affected department identifies revisions to programs, processes, and documents required as a result of the design change. These revisions are made in accordance with plant procedures ensuring that plant SSCs are installed and operated in accordance with established design requirements and plant documentation. CM submits the PC/M package to the Facility Review Group (FRG), a multi-discipline group, for review and recommendation to the Plant General Manager (PGM) for approval. Upon approval by the PGM, the design change may be implemented. The Company Nuclear Review Board (CNRB) also provides an independent review of the design modifications to verify that the change did not constitute a USQ. Both the FRG and the CNRB reviews are Technical Specification requirements for St. Lucie.

2.1.2.3.1 Changes to PC/M After a PC/M has been issued for implementation and prior to final closure, ifa minor technical change or a documentation change to the PC/M is required, it is evaluated for disposition via the Change Request Notification (CRN) process. The engineering design organization reviews and resolves CRNs. Special generic updates to P/CMs are issued periodically. These PC/Ms are used to make equivalent replacements, change drawings, or execute minor configuration changes bounded by the scope of the original PC/M. These changes are processed as CRNs to the PC/M.

Ifthe change to the PC/M affects the safety evaluation, design criteria or original PC/M intent, then the PC/M is revised and the revisions to the PC/M receive the same review as the original PC/M.

2.1.2.3.2 PC/M Closure The Configuration Management (CM) department is the plant organization specifically tasked to supervise processes affecting configuration management and to ensure that the plant design and configuration are maintained accurately. The implementing department provides the implementing documents to CM as a turnover package. Formal verification of completion by CM entails verifying the completeness of open items from the implementing department, satisfactory completion of PC/M requirements including implementation instructions, tests and inspections, and compliance to governing procedures.

2.1.2.4 Documentation Updates Drawings required by Operations for the safe operation of the plant are defined by St. Lucie and Engineering procedures as Plant Operational Drawings (POD). As part of the PC/M closure process, these drawings and appropriate plant operating procedures are updated and placed in the Control Room prior to the Operations Department accepting the modification. The remaining documentation is updated in accordance with approved procedures,

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 15 2.1.3 Non-PC/M Process Non-PC/M processes allow administrative document update, equivalent replacements, and minor, non-nuclear safety-related design changes. The non-PC/M processes, including document updates and equivalent item replacements are intended for changes that are outside of the scope of 10 CFR 50.59, but a screening is conducted as appropriate to ensure the change does not require a 10 CFR 50.59 safety evaluation or Technical Specification change. The non-PC/M process controls temporary system alterations, equivalent replacements, minor scope changes, drawing change requests, and design document updates. These activities are controlled under Engineering and Plant procedures.

The Temporary System Alterations (TSA) process, AP 0010124, "Control and Use of Temporary System Alterations," maintains plant and design configuration control for non permanent changes to plant SSCs, while ensuring the applicable technical and administrative reviews and approvals are obtained. The TSA process ensures that personnel are aware of temporary plant configuration and that TSAs made to plant equipment do not unacceptably degrade the original design intent.

To maintain plant configuration control, the affected drawings are stamped in the work control center and control room and identified as affected by TSA.

The Item Equivalency Evaluation (IEE) process is used to perform equivalent item replacements requiring implementation that are within the bounds of a maintenance activity. An IEE is required ifa replacement item is not a like-for-like (identical) replacement for the original item. IEEs determine and document the acceptability of non-identical replacement items by evaluating form, fit and function. Implementation of the replacement item must fall into the maintenance arena (no modifications or construction). Ifnot, the appropriate design change vehicle is required to replace or augment the IEEs.

The Drawing Change Request (DCR) process is used to perform administrative changes to design documents for which there is an approved basis (no physical change or design change). Typical uses of DCRs are administrative changes to controlled documents, or changes in controlled documents substantiated by IEEs.

The Non-Nuclear Safety (NNS) modification process is used to perform minor NNS changes in support of maintenance activities which have a technical relationship or interface with the plant, and do not affect PODs or plant operating procedures. Facilities such as the water treatment plant or components such as air conditioners for the service building may be modified by this process. Typically the Condition Report process initiates these minor modifications.

Certain engineering specifications referred to as "Re-use," provide generic guidance, instructions and details which can be used for routine repetitive maintenance activities. These engineering specifications are supported by 10 CFR 50.59 safety evaluations, ifrequired. Each specification identifies the scope of activities it addresses. The appropriate specifications can be invoked within plant documents e.g., PWOs, for routine maintenance activities, Ifplant Maintenance determines

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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 16 during the implementation of a PWO that the maintenance activity may not be covered by the engineering specification and a change is required, an engineering review is provided.

Engineering reviews the maintenance activity and either approves or disapproves the change requested. Ifthe design change process is entered, e.g., an REA may be issued.

2.1.4 Types of Evaluations When design changes are desired/required the safety evaluation process is utilized to demonstrate that safety is maintained in the modification process. There are different types of evaluations performed by Engineering and other qualified personnel for different aspects of plant support.

These include general evaluations (studies, feasibility assessments), operability assessment, nonconformance assessment and resolution, condition assessment and resolution, proposed license amendments (10 CFR 50.90), probabilistic safety assessment, etc.

2.1.4.1 Stand-Alone Safety Evaluations Stand-alone safety evaluations (SASE) are used for meeting 10 CFR 50.59 requirements in which no hardware change is involved, thus an EP may not be required. The SASEs include a concise description of the proposed change, why the change is necessary, the effect on plant operation and safety, any changes in design or operating practices or philosophy, any restrictions on plant operations, whether a USQ exists or Technical Specification change is involved. The evaluation also describes the licensing requirements including the information normally contained in the design section of an EP and the actions required including short-term and long-term actions by the plant and/or engineering departments. The affected documents are identified and the appropriate change package is attached to the evaluation, including a list of applicable documents used in preparation of the evaluation. These evaluations are submitted to the FRG for review and recommendation for approval by the PGM. Any action items listed in the evaluation are tracked using the Plant Management Action Item (PMAI) process.

2.2 Configuration Control Processes 2.2.1 General Description The configuration control processes integrate all plant activities. These processes accomplish the CM process objectives described in Section 1.0 above by defining and documenting the requirements for maintaining the plant's configuration (design, physical plant, and procedures) consistent with the plant design bases.

The governing procedures for these processes are Engineering and Plant QIs and other applicable plant procedures some of which are listed below:

ENG QI 1.0 Design Control ENG QI 1.1 Engineering Packages (EPs)

ENG QI 1.2 Minor Engineering Package (MEP)

l St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 17 ENG QI 1.3 Drawing Change Requests (DCRs)

ENG QI 1.4 Change Request Notices (CRNs)

ENG QI 2.0 Engineering Evaluations ENG QI 2.1 10 CFR 50.59 Screening/Evaluation ENG QI 2.4 Non Conformance Reports (NCRs)

ENG QI 2.5 Condition Reports ENG QI 3.0 Quality Assurance Records ENG QI 3.1 Controlled Document Distribution ENG QI 3.2 Drawing Control ENG QI 3.3 Vendor Technical Manual Control ENG QI 3.4 FSAR Updating ENG QI 3.5 Design Basis Document (DBD) Updating ENG QI 3.6 Total Equipment Data Base (TEDB)

ENG QI 4.2 Procurement Engineering Control ENG QI 6.7 FSAR Reviews QI 3-PR/PSL-1 Design Control QI 6-PR/PSL-1 Document Control ADM-08.04 Root Cause Evaluation AP 0006129 PMAI Corrective Action Tracking System QI-5-PS L-1 Preparation, Revision, Review/Approval of Procedures AP 0010148 Temporary Changes to Procedures OP 0010122 In Plant Equipment Clearance Orders ADM-17.10 Processing Safety Evaluations ADM-17.03 Operating Experience Feedback AP 0006130 Condition Reports ADM-0010432 Control Of Plant Work Orders ADM-78.01 Post Maintenance Testing 2.2.2 Integration of Configuration Control Processes Configuration control processes establish the authorities and responsibilities for integrating plant activities which could affect the plant configuration. They ensure that the integrity of the design bases is maintained by requiring changes that may impact the design bases to be reviewed by the proper organization. These changes include physical changes to the plant that may originate through the modification process, or daily operational and maintenance tasks, The design bases, in turn, must be accurately reflected in the engineering, licensing, operating, maintenance, testing, training, and quality assurance documents for the plant with consistency maintained among these documents.

Design, Physical, and Operational Plant Configuration Change Processes The design change PC/M package preparation activities are performed by Engineering and must account for affected design and administrative documents. Design integration reviews are performed to ensure that "open" plant changes in progress do not conflict or compete with the

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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 18 new design change. Relational databases to track change documents and other design integration tools are available to facilitate this design integration review. Walkdowns are performed to validate the design change packages against field conditions. As-building of affected documents is automatically triggered by the relational document control databases upon completion of a design change package.

Part of the design change activity is the procurement control process that provides a mechanism to purchase initial parts and services for PC/Ms, and subsequent stock replacement parts.

Plant work is controlled and documented with the Plant Work Order (PWO) process. This process is a mechanism to provide the administrative requirements and work controls for plant work activities on Safety-Related (SR), Quality-Related (QR), and NNS SSCs. The PWO is the primary implementing document for PC/Ms. PWOs become the plant's historical work records.

When work is complete, the information provided in the PWOs is entered into FPL's PassPort database for easy retrieval.

Changes to the plant's operational configuration that are not controlled by procedures, are controHed and documented with the clearance process, OP-0010122, "In-Plant Equipment Clearance Orders." This process tracks the removal and restoration of equipment out of service.

The clearance process provides administrative controls that allow for the isolation of components for maintenance activities, or for the safety of plant personnel and equipment. Each clearance is checked against Technical Specification requirements, and risk significance is reviewed prior to the issuance of the clearance. Multiple reviews and checks are performed to ensure that the components are isolated and restored correctly to operable'status.

Design changes are required to be reviewed by interfacing organization personnel knowledgeable in plant processes and configuration, and by system engineers knowledgeable about the affected SSCs, Operations, maintenance and testing personnel perform interface reviews for FRG-affected procedures. A final review by the FRG is performed for safety impacts.

During implementation, any necessary field changes must first be documented and approved by Engineering, via the CRN process, before they are implemented. Post-modification testing requirements are part of the design change package and test results become part of the design change permanent records. Implementation documents such as temporary procedures, test results, etc., are submitted to CM for review as an implementation turnover package (ITOP).

This package is then used to prepare a system acceptance turnover sheet (SATS) to document turnover of the modified equipment/system to Operations. This system acceptance and turnover process which is used as part of the final design change implementation, requires mandatory approval from all affected plant departments.

High priority affected documents (operating procedures, plant operating drawings (POD), etc.)

receive first priority for update upon notification of design change implementation completion.

These documents must be updated and available as part of the final SATS. Training is conducted on design changes and operations procedures prior to final system acceptance, as appropriate.

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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 19 Qe 2.2.2.2 Procedure Change Process This process provides a mechanism to make procedure changes in a controlled manner.

Procedure changes may be either permanent or temporary. New procedures and changes to existing procedures require as a minimum, a 10 CFR 50.59 screening to determine ifthe new procedure or change is consistent with the FSAR, Technical Specifications or whether a USQ exists. Ifa 10 CFR 50.59 safety evaluation is prepared for new procedures or changes to an existing procedure, the changes are reviewed by the FRG. Procedures required to be reviewed by Technical Specification are reviewed by the FRG. New procedures are validated prior to use and revised procedures may be validated ifthe change is extensive in nature.

2.2.2.3 Document Control Process The document control process governs the preparation, issuance, and revision of documents such as instructions, procedures, and drawings which describe activities affecting quality. Document control processes ensure that documents including changes, are reviewed,'pproved for release by authorized personnel, and distributed to the responsible individuals/organizations for implementation and use as appropriate. Databases or hard copy logs that identify the most recent approved revision of documents are maintained. Controls that prevent the inadvertent use of e obsolete or superseded documents are provided. Document control process responsibilities include identifying, storing, updating, and retrieving appropriate documents throughout the life of the plant. The FSAR and DBD updating processes are described in section 5.1 and 5.2, respectively which addresses the implementation process for 10 CFR 50.71(e) requirements.

2.2.2.4 Quality Assurance Records Records are maintained to provide documentary evidence of the performance of activities affecting quality including design activities, 2.2.3 Discrepancy Documentation and Commitment Tracking Processes 2.2.3.1 Condition Report (CR) Process The CR process allows personnel to document, evaluate, analyze and correct conditions of concern. CRs are initiated for nonconformances and other events or conditions that may appear to be adverse to the safe and orderly conduct of plant operations. The CR process is not intended to replace or duplicate the functions of other St. Lucie programs which provide for identification, disposition and trending of adverse conditions, e.g., PWOs, REAs . The CR originator delivers the CR to the Nuclear Plant Supervisor (NPS). The NPS reviews each CR for operability and reportability concerns, logs it and forwards it to the PGM. The PGM screens each CR and assigns the investigating group responsible for the investigation, analysis and determination of corrective actions ifrequired. The Condition Report Oversight Group (CROG) reviews daily the CRs generated in order to verify reportability, outage significance, and classification of condition.

~ I . Il St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 20 The dispositioning Department Head determines ifthe root cause is adequately identified and if the corrective actions are adequate to prevent recurrence. Closing a CR requires that the PGM accepts satisfactorily completed evaluations. Ifthe corrective actions can not completed within the designated time frame, the remaining open items are transferred to the Plant Management Action Item (PMAI) or other tracking system.

2.2.3.2 Nonconformance Report (NCR) Process The NCR is considered a subset of the CR process which provides a means for capturing potential or real concerns with respect to the plant, and allows for their correction using other existing and integral plant processes, e.g., the PC/M process, PWO. Nonconformances have historically been addressed using an NCR. The NCR provides a mechanism for documenting the identification of a nonconformance, getting appropriate reviews, and resolving the nonconformance or taking corrective actions. This process is used to collect data for future improvements. Engineering evaluates the NCR and provides an appropriate disposition (i.e., corrective action, ifany) to the NCR for implementation.

2.2.3.3 Plant Management Action Item (PMAI) Process The PMAI process provides a method for controlling and scheduling corrective actions to meet internal commitments and ensures follow-up and closure of long-term corrective actions generated by the CR process. The PMAI is a tracking system which follows the corrective actions to completion and gives the items visibilitywithin the plant organization.

2.2.3.4 Commitment Tracking Process A commitment tracking process is in place to provide a method of identifying and tracking commitments. Sources of commitments include Licensee Event Reports, Notice of Violation responses, NRC requirements, INPO or internal audit findings, and NRC safety evaluation reports. The major objective of the process is to assure completion of the commitments in a timely manner.

3.0 Implementation of 10 CFR 50.59 The 10 CFR 50.59 safety evaluation provides assurance that the documented information used by the NRC, as a basis for licensing the facility, remains valid in light of the proposed changes. FPL uses NSAC 125 as guidance for safety evaluations and 10 CFR 50.59 screenings.

When proposed design changes are desired or required, the 10 CFR 50.59 process determines if the change requires a revision to Technical Specifications or ifthe change constitutes a USQ.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 21 3.1 Unreviewed Safety Question Determination Process Changes to the facility, procedures described in the FSAR, or a test or an experiment not described in the FSAR are screened for the need to determine ifa USQ exists. This section discusses the definitions of a change, the 10 CFR 50.59 screening process, and the preparation of a formal safety evaluation in accordance with 10 CFR 50.59.

3.1.1 Change to the Facility Changes to the facility include additions or deletions of SSCs described in the FSAR, modifications which affect design function or method of performing a function of an SSC, and temporary modifications.

3.1.2 Change to the Procedures FPL revises procedures as described in the FSAR without prior NRC approval provided the change does not involve a Technical Specification change or a USQ. Procedures that may require a 10 CFR 50.59 safety evaluation are those that are outlined, summarized or completely described in the FSAR.

3.1.3 Tests and Experiments Tests and experiments that require a 10 CFR 50.59 safety evaluation are those that are not discussed in the FSAR, but are within the scope of the FSAR, i.e., tests and experiments which could degrade the margins of safety during normal operations, anticipated transients, or that could degrade the adequacy of SSCs to prevent accidents or mitigate accident conditions. 10 CFR 50.59 safety evaluations are also required for changes to tests or experiments that are described in the FSAR.

3.2 10 CFR 50.59 Screening and Evaluation Process St. Lucie facility changes are evaluated for 10 CFR 50.59 applicability. This evaluation is referred to as a 10 CFR 50.59 screening. Formal safety evaluations are prepared for those changes that fall within the scope and intent of the 10 CFR 50.59 rule. The applicable procedures to the design change process contain guidance on screening criteria. The following criteria are used in screening activities for 10 CFR 50.59 applicability:

1. Does the change represent a change to the facility as described in the FSAR?

2. Does the change represent a change to procedures as described in the FSAR?

3. Is the change associated with a test or experiment not described in the FSAR?

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 22

4. Could the change affect Nuclear safety in a way not previously evaluated in the FSAR?

5. Does the change require a change to the Technical Specifications?

A positive response to any of questions 1 - 4 requires a formal 10 CFR 50.59 safety evaluation. A positive response to question 5 requires a license amendment in accordance with 10 CFR 50.90 prior to change implementation. Furthermore, a design change which requires a change to the Technical Specifications requires NRC review and approval of a license amendment prior to implementing the change.

3.3 The 10 CFR 50.59 Safety Evaluation For the purpose of a safety evaluation the 10 CFR 50.59 criteria have been expanded into seven questions for clarity:

1. Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the SAR?

2. Does the proposed activity increase the consequences of an accident previously evaluated in the SAR?

3. Does the proposed activity increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR?

4. Does the proposed activity increase the consequence of a malfunction of equipment important to safety previously evaluated in the SAR?

5. Does the proposed activity create the possibility of an accident of a different type than any previously evaluated in the SAR?

6. Does the proposed activity create the possibility of a different type of malfunction of equipment important to safety than previously evaluated in the SAR?

7. Does the proposed activity reduce the margin of safety as defined in the basis for any Technical Specifications?

Based upon the answers to these questions, a determination is made ifa USQ exists or a Technical Specification change is required. Ifeither exists, NRC review and approval of a license amendment is obtained prior to the change being implemented.

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The primary participants in the 10 CFR 50.59 safety evaluation preparation are as follows;

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Preparer: A properly trained and qualified preparer who is familiar with the 50.59

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process develops an understanding of the design basis of the plant and applicable

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 23 regulatory requirements. The design change under consideration typically falls within the preparer's field of expertise.

b. Emmy: A properly trained and qualified reviewer that has not been participating in the preparation of the 50.59 evaluation verifies the acceptability of the safety evaluation.

c. ZRQ: The FRG is composed of designees representing the functional areas of the plant and provides a multi-discipline review of the safety evaluations in accordance with 10 CFR 50.59 guidelines and concurs with the determination that a USQ does not exist and no Technical Specification change is required.

Additionally, the following after-the-fact reviews are conducted:

d. Q5RB: An offsite safety review board, the CNRB, composed of officers, managers, or specialist in design, operations, safety analysis or related activities, with extensive nuclear plant expertise is responsible for the review of evaluations prepared according to 10 CFR 50.59 guidelines. The purpose of the review is to confirm the change does not constitute a USQ and no Technical Specification change is required,

e. ~: Quality Assurance performs periodic audits of organizations involved in the 10 CFR 50.59 evaluation process, Audit results are documented and reviewed by management having responsibility in the specific areas audited. Corrective actions are documented and follow-up audits are performed, ifrequired.

The 10 CFR 50.59 safety evaluation process is applied to potential design, physical plant and procedural changes that may impact the plant license, design bases or the FSAR. However, there are changes to the plant which are not governed by the 10 CFR 50.59 requirements and are managed under St. Lucie procedures. These are referred to as non-10 CFR 50.59 activities as discussed below.

3.4 Non-10 CFR 50.59 Activities Plant activities, controlled under various administrative, engineering, operations and maintenance procedures, which are not normally considered design activities, but which could possibly affect the design basis are screened for 10 CFR 50.59 applicability. Each procedure controlling these activities contains a requirement to perform a 10 CFR 50.59 screening. These screenings are performed by qualified personnel trained on the 10 CFR 50.59 requirements.

4.0 Engineering Design and Configuration Control Process Training St. Lucie training includes indoctrination, technical and continuation training for engineering and supporting interface department personnel. This training provides instruction and on-the-job training required to ensure compliance with St. Lucie procedures/policies regarding plant

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 24 activities, including design. Formal training activities are governed by two Quality Instructions:

QI 2-PR/PSL-2, "Indoctrination and Training of St. Lucie Personnel," and QI 1-PR/PSL-11, "Site Services Organization." Training materials are developed from and maintained consistent with currently issued FPL and St. Lucie plant design documentation, procedures and policies. The St.

Lucie Training Department is on distribution for updated plant design bases documents, procedures and FPL policy notices. In addition, the St. Lucie Training Department is on distribution for relevant NRC and industry correspondence regarding power plant requirements and activities. New or revised training materials are developed which are current with the best industry practices.

Training associated with design and configuration control processes that implement 10 CFR 50.59, 10 CFR 50.71(e), and Appendix B to 10 CFR Part 50 are discussed in the following sections.

4.1 Engineering Support Personnel (ESP) Training and Qualification Initial indoctrination training for engineering support personnel includes a basic review of the plant QIs and procedures associated with the configuration control process, the plant modification process, the use of a 10 CFR 50.59 safety evaluation to modify the plant or identify an unreviewed safety question. Each of the reviews is part of the initial indoctrination training section of the INPO accredited ESP Training Program for all new engineering support personnel.

The ESP Qualification Guides provide individual qualification requirements. These qualification requirements are based on an evaluation of the individual's education, previous experience and technical training. The training required for each individual is specified by the department supervisor.

4.2 Shift Technical Advisor (STA) Training Program The STA is responsible for reviewing procedure changes for applicability to the requirements of 10 CFR 50.59. STAs are trained on the requirements of 10 CFR 50.59 and the requirements of the CR process.

4.3 Nuclear Plant Supervisor (NPS) Training Program The NPS is responsible for reviewing CRs for safety issues and operability concerns. The NPS training program contains specific lessons on plant changes in accordance with 10 CFR 50.59 and applying the plant's design basis to operations.

4.4 Facility Review Group (FRG) Training Program Training on 10 CFR 50.59 requirements is presented as part of the FRG initial and requalification training. It is designed to introduce FRG members and alternates to their responsibilities as related to FRG review of procedure and design changes.

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St. Lucic Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 25 5.0 10 CFR 50.71(e) Implementation Process The engineering design and configuration control processes described in the response to this request, detail the activities that ensure the accuracy of the changes which are made to the FSAR.

5.1 FSAR Updates The St. Lucie Unit 1 and Unit 2 FSARs were originally submitted to the Nuclear Regulatory Commission (NRC), as part of the operating license application, and are maintained and updated in accordance with the requirements of 10 CFR 50.71(e)(3)(l). Amendments reflect FSAR changes up to a maximum of 6 months prior to the date of filing. Special FSAR amendments may be submitted to the NRC more frequently than the minimum required by 10 CFR 50.71(e). ENG QI 3.4, "FSAR Updating," provides the instructions for updating the FSAR in accordance with 10 CFR 50.71(e),

FSAR amendments consider the following categories of items implemented, revised, or identified since the last amendment:

~ Changes made in the facility/procedures as described in the FSAR

~ Safety analyses performed in support of license amendments

~ Safety evaluations performed per 10 CFR 50.59

~ Safety evaluations performed at the request of the NRC

~ New or modified NRC requirements

~ Changes made to drawings contained in the FSAR

~ Commitments in FPL'orrespondence to the NRC

~ Corrections to the FSAR

~ Changes to design bases or design criteria identified in the FSAR

~ Changes to quality, procedural, test, inspection, or other criteria cited in the FSAR

~ Abandoned equipment noted in the FSAR Engineering design changes are procedurally required to address the impact on the FSAR as part of the engineering design change process, and to include proposed changes to the FSAR with the issuance of each design change package, as applicable. This process provides a comprehensive mechanism for capturing changes to the FSAR and for their incorporation into the design bases.

Changes to the facility may affect the FSAR descriptions and associated accident analyses.

Engineering design outputs, e.g., EPs, stand-alone Safety Evaluations, which impact the FSAR include an FSAR Change Package (FCP) as an attachment to the document. FCPs are reviewed, and ifadequate, are incorporated into the next amendment of the FSAR following the implementation of the change,

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Plant design changes implemented during the update period and all 10 CFR 50.59 safety

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evaluations are reviewed for impact on the FSAR. When a change is proposed, the organization

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initiating the change determines whether the change willresult in the revision of any documents.

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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 26 Changes to licensing documents can result from design changes, as-found conditions, evaluations, procedure revisions, and NRC requirements.

FPL responses to NRC Information requests, IE Bulletins/Circulars/Information Notices, FPL correspondence to and from the NRC, etc. may affect the FSAR. Incoming and outgoing NRC correspondence is reviewed for impact on the FSAR. Ifit is determined that an FSAR change is required, an FCP is prepared along with any necessary supporting evaluations.

The intent of the NRC correspondence review is to identify and document NRC requirements that impact the design, operation, or maintenance of the facility, regardless of the level of detail associated with the requirement. Commitments in NRC correspondence to implement physical changes to the facility are not incorporated into the next FSAR revision. Physical changes are only incorporated after implementation and turnover for use.

Stand-alone safety evaluations that require FSAR changes have an FCP attached. Engineering reviews stand-alone safety evaluations forwarded by other departments and/or contractors for potential FSAR impact. FCPs are developed as necessary.

The FSAR user comment form, which is proceduralized, permits the FSAR user to identify actual or perceived discrepancies in the content or accuracy of the FSAR. FSAR user comments concerning the content of the FSAR are forwarded to Engineering for review and incorporation into the FSAR. The FSAR user comment is reviewed to determine ifenough information is provided. In order to be processed for inclusion in an FSAR amendment, the user comment must meet one of the following criteria:

I) Editorial, such that back-up documentation is not required.

2) Providing sufficient bases, e.g., safety evaluation, PC/M reference, or equivalency justification to show that the desired change has been evaluated and/or analyzed.

Ifthe criteria stated above is not met, the deficiencies are documented on the user comment form and returned to the originator. Ifthe user comment is adequate, it is forwarded to the cognizant discipline for review and disposition. (Note that user comm'ents are typically forwarded to Engineering as part of a CR.) Ifa valid error or concern is identified, the comment process will initiate those activities required to correct/update the FSAR prior to submittal to the NRC. A completed FSAR amendment is reviewed for accuracy and completeness by the appropriate plant departments prior to submittal to the NRC.

5.2 Design Basis Document Updates The Design Basis Documents (DBD) for St. Lucie have been issued. Some change packages to this document have been generated but have not been incorporated into the document. Periodic updates are planned on a schedule consistent with the FSAR and willfollow the process for updating in accordance with ENG QI 3.5, "Design Basis Document (DBD) Updating."

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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 27 DBD updates consider the following categories of items implemented, revised, or identified since the last revision;

~ Changes to design bases or design criteria identified in the DBD

~ Changes made in the systems/components described in the DBD

~ Safety evaluations performed per 10 CFR 50.59

~ Safety evaluations performed in support of license amendments

~ Safety evaluations performed at the request of the NRC

~ New or modified NRC requirements

~ Commitments in FPL correspondence to the NRC

~ Changes made to figures contained in the DBD

~ Corrections to the DBD Changes to the facility may affect the DBD descriptions. These changes may result from design changes issued by Engineering, Contractors, or other St. Lucie departments. Engineering design outputs, e.g., EPs, MEPs, DCRs, evaluations, which impact the DBD include a DBD Change Package (DCP) as an attachment to the document. The level of detail and format provided in DBD updates are consistent with the level of detail and format contained in the original DBD.

The review of NRC/FPL correspondence and stand-alone safety evaluations for impact on the DBD is the same as the method used to review the FSAR updates discussed in Section 5.1. DBD user comment forms are proceduralized and used similar to FSAR user comment forms (see Section 5.1) to identify actual or perceived discrepancies in content or accuracy of the DBD.

Prior to issuing a DBD update, a review of the revised DBD is coordinated with applicable departments and organizations.

6.0 10 CFR 50, Appendix B Implementation Process The FPL Quality Assurance (QA) Program is described in the FPL Topical QA Report (TQAR) and is structured to be in compliance with the requirements of Appendix B to 10 CFR Part 50.

The TQAR delineates the QA Program requirements and summarizes the FPL approach to activities related to materials, parts, components, systems and services included in the QA Program. The TQAR states that a QA Program be established for design-related activities. More specifically, the TQAR provides the general guidance that the design control program must ensure that the design is defined, controlled and verified, that applicable design inputs are specified and correctly translated into design output documents; that design interfaces are identified and controlled; that design adequacy is verified by persons other than those who designed the item; and that design changes, including field changes, are governed by control measures commensurate with those applied to the original design.

The general design requirements described in the TQAR are implemented through QIs. For each applicable criterion in 10 CFR 50, Appendix B, there is a corresponding Engineering and Plant QI or series of QIs that comply with the criterion in Appendix B.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 28 The design records are developed to provide evidence that the design process and design verification were performed in accordance with the requirements of St. Lucie's QA Program.

QIs are developed as required by each of the implementing departments that describe the measures to be used to implement the quality assurance requirements. QIs describe actions and responsibilities to be performed within a department or organization and address the requirements of the appropriate TQAR requirements.

QA performs periodic audits of the configuration control processes. The FPL TQAR requires that audits be regularly scheduled for on-going activities, and the scope of those audits include the preparation, review, approval and control of the FSAR, designs, specifications, procurement documents, instructions, procedures, and drawings. QIs require a biennial functional area audit of design and configuration control to be performed by the QA organization. Design control is also audited during audits of other functional areas such as fire protection, fuels, refueling operations, environmental protection, and protection and control. Activityaudits of specific design and configuration control issues may be performed at any time. In addition, reviews of safety-related and non safety related SSCs have been performed as part of audits and technical reviews. Also, St. Lucie procedures require that procedures and instructions (including those related to design and configuration control) be reviewed by QA for compliance with the TQAR, and other applicable industry standards and requirements.

TQR 18.0, "Audits," requires that QA Audit Findings be issued to the responsible management of the audited organization, who are required to correct the deficiencies identified in the audit report and take actions to prevent their recurrence. The status of corrective actions are tracked by QA until the corrective actions have been accomplished and verified by QA.

St. Lucic Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Pago 29

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

The following outline provides the organization of the response to request [b]:

1.0 Introduction 2.0 Programs Affecting Design Basis 2.1 Commitment to Excellence Program 2.1.1 Operating Procedure Upgrade Project 2.1.2 Technical Specification Improvement Project 2.2 Design Basis Program 2.2.1 Design Basis Document Development 2.2.2 Design Basis Document Use and Maintenance 3,0 Recent Design-Related Projects 3.1 Plant Procedure Improvement Program 3.2 Containment Penetration Review 3.3 Instrument Setpoint Verification 4.0 Functional Review and Verification of Design Basis Translation into Procedures 4.1 FSAR Procedural Consistency Review 4.2 Annunciator Summary Review 4.3 Independent Safety Engineering Group 4.4 Inservice Inspection and Testing Programs 4.5 Surveillance Testing 4.6 NRC Inspections 4.6.1 Emergency Operating Procedure Team Inspection 4.6.2 Operational Safety Team Inspection 4.6.3 Maintenance Team Inspection 4.6.4 Equipment Environmental Qualification Inspection 4.6,5 Conclusion from NRC Inspection Efforts 4.7 FPL Audits, Self-Assessments and Findings 4.7.1 Audit Findings and Strengths 4.7.2 1996 Plant Self-Assessment 5.0 Conclusion

b' St. Lucio Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 30 1.0 Introduction FPL has reviewed the processes used to license, design, operate and test the St. Lucie Plants.

This review has indicated that the processes will maintain the design basis and assure that operating, maintenance and testing procedures preserve the design basis. This includes the recently completed FSAR Procedural Consistency Review and the self-assessment performed in preparation for the recent NRC A/E Inspection. The rationale for concluding that the St. Lucie design basis requirements are translated into plant procedures is based upon the following:

1. The design, procedures and plant modifications have been under procedural control since initial operation to assure that changes to the procedures, the design or to the FSAR are consistent. The St. Lucie programs and initiatives which have aided in assuring this consistency were the Commitment to Excellence Program, plant procedure improvement program, and the containment penetration review.

2. The startup test procedures verified that the plant operated as designed and met the design bases for plant operation.

3. The Design Basis Program helped to ensure that the design bases was correctly incorporated into the plant's operating procedures.

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4. Current and past plant programs, such as the plant procedure improvement program and the instrument setpoint verification, confirm the acceptability of plant procedures.

5. Self-assessments, FPL audits, and NRC inspections have evaluated through sampling programs that the processes for maintaining procedural consistency with the design bases are acceptable and provide reasonable assurance of their effectiveness.

The original design process produced design documents such as calculations, drawings, specifications, evaluations, and analyses necessary to support initial construction, testing, operation and licensing of the plant. The design documents produced by this process were used as the bases for system startup and acceptance testing. The design and construction process included various quality assurance audits, quality control inspections, and documentation requirements to assure consistency between construction and design.

The FSAR was created during the plant design process and was submitted to the NRC for review and approval. The FSAR includes descriptions of the plant's design and operating practices required to show compliance with NRC regulations. NRC review of the original FSAR resulted in additional amendments prior to plant operation. The NRC Safety Evaluation Report documents this review and the NRC staff's findings that the plant design and operating practices were acceptable, Technical Specifications were developed (in conjunction with the NRC review) to identify functional requirements, controlling parameters, and surveillance and testing requirements that are significant to the safe operation of the plant. The maintenance procedures

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Pago 31 were developed from the vendor technical manuals, and A/E and NSSS supplier documents which contained the design bases requirements for SSCs.

These licensing and design reviews conducted in conjunction with the initial startup testing of the plant, confirmed that the design bases requirements were appropriately translated into plant procedures prior to initial operation. The initial startup test program was developed based on design basis requirements identified by the NSSS supplier and the A/E. The completion of the startup testing verified that the plant's systems operated as designed and that the plant was configured in accordance with the design basis. The startup test report was submitted to the NRC prior to commercial operation. These test procedures provided a basis for the operating and testing procedures.

The Commitment to Excellence Program (CEP) included projects to upgrade operating procedures and to establish better consistency in plant operations thro'ugh the use of a standard format for Technical Specification. The Design Basis Document development program consolidated information contained in reference design bases sources and compared plant procedures to design basis requirements.

The plant procedure improvement program, containment penetration review, and instrument setpoint verification project are examples of efforts that help to confirm that current plant procedures correctly implement the plant design bases. In addition, the numerous FPL plant reviews, self-assessments and audits, and NRC inspections, have all contributed to ensuring that the plant design bases, are correctly reflected in the plant's procedures.

These projects, programs, initiatives and processes collectively support the rationale for concluding with reasonable assurance that the operations, maintenance and testing procedures at St. Lucie are consistent with the plant's design, and that discrepancies between design documents, the FSAR and the plant are identified and resolved.

2.0 Programs Affecting Design Basis The following projects support the rationale that the procedures reflect the design basis.

2.1 Commitment to Excellence Program The St. Lucie "Commitment to Excellence Program" was modeled after the Turkey Point "Performance Enhancement Program (PEP)" and was developed to meet four specific goals:

~ Continued safe and reliable plant operation.

~ Improved plant and site material condition.

~ Increased emphasis on quality performance in systems, controls, and personnel.

~ Continued responsiveness to regulatory requirements and corporate goals.

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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 32 Various improvement elements were brought under one broad program to enhance coordination efforts, better utilize resources, and provide the proper control to ensure accomplishment of the program goals. The CEP program was specific to St. Lucie Plant and incorporated, where appropriate, the results and lessons learned from the PEP implemented at FPL's Turkey Point Plant. Two of the CEP projects that relate to procedure improvements are:

~ CEP Project 3 - Operating Procedure Upgrade Project

~ CEP Project 9 - Technical Specification Improvement Project 2.1.1 Operating Procedure Upgrade Project CEP Project 3 was the St. Lucie project to develop and implement a program to upgrade and maintain plant operating procedures. The program included the establishment of operating procedure upgrade resources, revision of emergency operating procedures (EOP), establishment of consistent procedure formats, revision and upgrade of plant operating procedures, training of personnel on upgraded procedures, and implementation of the revised procedures. The plant emergency operating procedures were revised in accordance with CEN-152, the generic EOP guidelines for Combustion Engineering plants, and NUREG 0737 commitments. The review and upgrade of the plant operating procedures addressed format, technical accuracy, Technical Specifications, source and reference documents, acceptance criteria, and applicable human factors criteria.

2.1.2 Technical Specification Improvement Project CEP Project 9 was the St. Lucie project to develop and implement a program for tracking proposed Technical Specification changes and to provide uniformity between Unit 1 and Unit 2 Technical Specifications. The project consisted of a review and validation of the differences between the Unit 1 and the Unit 2 Technical Specifications and the submittal of changes to the NRC.

2.2 Design Basis Program 2.2.1 Design Basis Document Development The Design Basis Program involved the development of a set of unit-specific system-level documents, referred to as design basis documents (DBD), that contain a roadmap to the reference regulations, codes, standards, calculations, analyses, specifications, etc., that form the basis for system design, testing and operation. Each DBD was reviewed against plant drawings in an effort to ensure consistency with the design bases and the plant configuration and operating procedures.

The DBDs explain system design and provide a definition of the bases for the design, component design constraints, and design features. The results of the Design Basis Program support the conclusion that the operating, testing, and maintenance procedures are consistent with the plant's design basis.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 33 2.2.2 Design Basis Document Vse and Maintenance The DBDs provide a means for facilitating the search and retrieval of design bases source documents, In this regard, the DBDs provide a reference for future procedure revisions and therefore willhelp maintain consistency between the design bases and plant procedures. The DBDs also provide a useful tool for design engineering to support plant modifications and operability evaluations.

Maintenance of the DBDs is important to ensuring design and configuration control of the plant.

The DBDs were intended to be updated at approximately the same frequency as the FSAR. A Condition Report (CR) was issued in 1996 against the DBD program concerning the untimely incorporation of DBD updates. The proposed corrective actions to this CR includes having the DBD updates incorporated on a schedule consistent with updating the FSAR.

3.0 Recent Design-Related Projects 3.1 Plant Procedure Improvement Program The St. Lucie Plant Procedure Improvement Program is focused on regulatory performance improvement. The scope of the project is to identify and improve operation and maintenance activities by upgraded task instructions. Included in the upgrade process is review of the design basis documents to ensure regulatory commitment compliance.

The Procedure Improvement Program includes operations, maintenance, and administrative procedures. The initial step included centralizing procedure writers from Operations, Electrical Maintenance, Instrument and Control Maintenance, Mechanical Maintenance, and Administrative Departments under a procedure development supervisor. This new group provides a single point of contact for procedural issues. The goal of this new group is to take the current set of plant procedures and identify tasks for which procedures do not exist, and identify existing procedures that need to be upgraded. These procedures will be placed into a standardized format and use industry-proven language techniques to give the end-user technically correct, first time implementable procedures.

The improved process includes the following fundamental attributes:

~ Define the scope of a technical subcommittee

~ Review procedures for critical characteristics and attributes

~ Include industry standard methods for procedure validation

~ Validate by end-user to ensure the procedure is sound and implementable

~ Establish a site procedure writer's guide which incorporates INPO and industry approved writing standards Another fundamental attribute of the improved process is to review applicable design bases documents that govern the activities and control the tasks described in each procedure. These

St. Lucio Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 34 bases documents include, but are not limited to, St. Lucie Technical Specifications, FSAR, Off-site Dose Calculation Manual, NRC correspondence, ABB-CE bulletins, calculations, analyses, evaluations, specifications and plant drawings.

3.2 Containment Penetration Review In 1991, an evaluation, "Technical Assessment of Containment Penetration Boundaries," was issued for each unit, to address plant events related to the control of containment isolation boundaries. The objectives of these evaluations were to:

1) perform a technical assessment of each mechanical containment penetration to determine those components whose integrity or operational position were vital to containment integrity; and

2) provide sketches identifying containment boundary components to assist plant personnel involved in operations and maintenance activities.

The evaluations establish, for each individual penetration, the containment penetration boundary

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including components that are vital to containment integrity.

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In preparing the evaluations, each penetration configuration was evaluated to:

1) confirm that the intent of the 10 CFR 50 Appendix A, General Design Criteria (GDC),

requiring a double barrier is satisfied for each line penetrating the containment; and

2) identify those components that form the containment penetration boundary, including those components that can be manipulated (e.g., valves opened, caps removed, or piping blind flanges removed) such that the double barrier criterion could be violated.

In each case, appropriate design and construction documentation was consulted and, where possible, the system components were walked-down to generate and confirm the containment penetration sketches.

The Unit 1 evaluation concluded that the isolation system meets the intent of the 10 CFR 50 Appendix A, General Design Criteria, in that a double barrier is provided in each line penetrating the containment. However, certain containment penetration valving did not conform in every detail to the requirements of GDC 54 through GDC 57, since the GDC were published after the Construction Permit was issued, and hence, were not available as a guide.

The Unit 2 evaluation concluded that the isolation system meets the requirements of the 10 CFR 50 Appendix A, General Design Criteria (exceptions taken to GDC provisions are discussed in Section 6.2.4.3 of the St Lucie Unit 2 UFSAR) in that a double barrier is provided in each line penetrating the containment.

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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 35 3.3 Instrument Setpoint Verification Between 1991 and 1993, FPL performed an instrument setpoint review for the St. Lucie Plant to verify and document current setpoints. Approximately 250 safety-related and important-to-safety instrument loops were evaluated with approximately 150 loops requiring documentation or recalculations. Plant operating and maintenance procedures were compared to the current field conditions and design documents to verify consistency. There were no design basis inconsistences found which were of a magnitude which required NRC notification. This program provides a basis to conclude that the current plant setpoints are correctly translated into plant procedures and that the design basis is available.

The original St. Lucie instrument setpoints were furnished by the A/E, the NSSS designer, or other equipment vendors, and have been supplemented through the years by engineering packages, which were typically produced by the A/E. Prior to the instrument setpoint verification effort, control of instrument setpoint change was governed by a separate process that involved plant and engineering procedures. In addition, it was an accepted practice at the time that some of the setpoints could be changed by the plant based on A/E or vendor input with limited documentation. Historically, the setpoint bases have been maintained in various documents including drawings, vendor manuals, and maintenance procedures, and the calibration/scaling values (typically calculated by maintenance) were maintained in the calibration procedure.

The setpoint verification effort (in addition to documenting or recalculating instrument setpoints) changed the old process by incorporating instrument setpoint changes into the configuration management process. The process change converted the setpoint list into an engineering drawing and placed instrument setpoints under engineering control in the configuration management process. Plant procedures were updated as appropriate to reflect the instrument setpoint bases described in the design drawings.

4,0 Functional Review and Verification of Design Basis Translation into Procedures Since initial startup, there have been numerous Quality Assurance and Engineering audits, self-assessments, NRC inspections, and third party reviews which have verified consistency between the plant, and the operations, maintenance, and testing procedures in use at the time of the inspection. These audits and inspections help provide reasonable assurance that the design and configuration control process has incorporated design basis changes resulting from plant modifications into plant procedures.

The engineering design and configuration management processes provide the necessary controls to ensure that the St. Lucie Plant design bases requirements are accurately translated into plant operating, maintenance, and testing procedures. For example, there have been several major self-assessment and improvement projects performed by FPL which have served to improve the St.

Lucie conflguration control processes. The effectiveness of these initiatives have been confirmed through formal NRC inspections and FPL audits of the engineering processes, design basis and procedures. Some of these improvement projects and inspection results are provided in support

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St. Lucio Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 36 of our rationale for concluding that the St. Lucie Plant design bases requirements are properly translated into operating, maintenance, and testing procedures:

FSAR Procedural Consistency Review Annunciator Summary Review Independent Safety Engineering Group Inservice Inspection Testing Program Surveillance Testing NRC Inspections QA Audits, Self-Assessments and Findings 4.1 FSAR Procedural Consistency Review During 1995 and 1996, NRC inspections found that plant operators were using procedures which had process steps not found in the FSAR and that the normal process steps found in the FSAR were not being followed. NRC Inspection Report 96-03 documented that licensed operators were diluting the reactor coolant system by directly injecting water into the charging header while the process steps described in the FSAR were to inject the water into the volume control tank to allow mixing. As a result of this event and other similar inconsistencies, FPL initiated a comprehensive FSAR review to improve consistency between operating procedures and the Oi FSARs. This effort was completed in December 1996.

The review of the St. Lucie Unit 1 FSAR identified 219 plant procedures as having some FSAR tie. The review of the St. Lucie Unit 2 FSAR identified 236 procedures as having some FSAR tie.

There were no plant modifications required as a result of these reviews.

The final conclusion of these FSAR reviews is pending the final results of some CRs, however, these CRs have had operability reviews and none are considered safety significant. The small number of FSAR inconsistencies requiring a CR indicates that there does not appear to be large difference between the FSARs and the plant operating procedures, 4.2 Annunciator Summary Review A plant annunciator summary review is currently in progress. As annunciators provide an early indication/announcement of possible degrading plant conditions, they are an important aid to operators in identifying appropriate procedural actions to take in response to these changing conditions. This program willreview selected annunciators for the control rooms, waste management panels, emergency diesel generators, steam generator blowdown building, boric acid concentrators, waste concentrators, water treatment plant, oxygen analyzer, and the liquid waste panels, as appropriate. This program reviews selected logic configurations, annunciator setpoints, procedural references, and corresponding operator actions found in the annunciator summary procedure for each unit against the design documents. The alarm setpoint is reviewed against the setpoint list which is a controlled engineering document. This project is approximately 10%

complete with over 200 annunciator setpoints, configurations and procedural responses reviewed

t4 St. Lucie Units l and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 37 and has not identified any inconsistencies which would be considered safety significant. This program continues to support the conclusion that the current plant procedures are consistent with the plant's design basis.

4.3 Independent Safety Engineering Group From 1983 to 1994, the Independent Safety Engineering Group (ISEG) performed technical reviews of plant safety systems. Concerns were forwarded to cognizant organizations along with recommendations to improve safety and performance. System walkdowns with plant and maintenance personnel were used to identify inconsistencies between the operating, testing, and maintenance procedures and the plant's configuration. Procedural adequacy was also verified as part of the effort and documented in reports. This program also supports the rationale to conclude that the operating, maintenance, and test procedures are consistent with the plant's design basis.

4,4 Inservice Inspection and Testing Programs Inservice inspections and tests are performed in accordance with ISI and IST procedures. The schedule of inspections and tests is in compliance with the requirements of the ASME Code. The procedural steps to perform a surveillance test includes the requirements to have plant operators in the control room review and approve performance of the test. The plant is placed in an acceptable configuration (consistent with the design basis) for the test, considering the current plant operating mode and Technical Specification requirements. After completion of the test, the test results are reviewed with the control room operators to assure components are operable prior to their return to service. CRs issued for components which do not meet acceptance criteria, ensure that a root cause of the failure is determined and resolved. The IST procedures ensure that plant operation remains within the design basis configuration during the testing.

4.5 Surveillance Testing Technical Specification surveillance tests and their schedules are performed in accordance with plant procedures. CRs issued for tests which do not meet acceptance criteria ensure the root cause of a failure is determined and resolved, Surveillance test procedures ensure that plant operation remains within the design basis configuration during the testing.

In accordance with the recommendations of Generic Letter 96-01, each reactor protection and safeguards surveillance test will be reviewed against the design basis for the system involved to assure that design basis requirements have been tested and that the surveillance test completely tests the entire actuation circuit and logic involved. The results of this review will be incorporated into the closure of Generic Letter 96-01, "Testing of Safety-Related Logic Circuits."

4 St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 38 4.6 NRC Inspections Selected NRC inspections and their results are discussed in the following sections. The results of these inspections were used to further improve the design and configuration control processes and also help to confirm that the St. Lucie design bases requirements have been maintained and properly translated into operating, maintenance, and test procedures. These inspections also confirm FPL findings that discrepancies found are not safety significant.

4.6.1 Emergency Operating Procedure Team Inspection In April 1988, the NRC conducted an Emergency Operating Procedure (EOP) Team Inspection at St. Lucie. The results of this inspection are documented in NRC Inspection Report 88-08 dated August 26, 1988.

This special announced inspection was conducted in the area of EOPs and included the implementation of vendor Generic Technical Guidelines, validation and verification, and training conducted on the EOPs. No unsafe operational conditions were identified. One deviation from a licensee commitment was identified in the report. Based on the NRC observations of the EOPs in use, the staff concluded that the EOPs were adequate for use by trained operators.

The staff did express concerns with a number of the findings made during the inspection which represent impediments to the effective implementation of the EOPs. Specific areas of concern identified included technical adequacy of the procedures (e,gvagueness, omissions, and deviations from suggested procedural steps), human factors (e.g., optical resolution of figures, not enough copies available to operators, and plant labeling deficiencies), and training weaknesses (e.g., technical basis presented in lesson plans occasionally disagreed with the EOPs and an operator was observed to follow the general guidance rather than the EOP).

Further inspector follow-up was performed for the noted discrepancies during NRC Inspection 90-32, which determined that the adequacy of corrective actions was sufficient to close the findings.

4.6.2 Operational Safety Team Inspection In April and May 1990, the NRC conducted an Operational Safety Team Inspection (OSTI) at St.

Lucie. The results of the inspection were documented in NRC Inspection Report 90-09.

This special announced inspection utilized a risk-based inspection guide and evaluated St Lucie s current level of performance in the area of plant operations. The inspection included an evaluation of the effectiveness of various plant groups including Operations, Surveillance/Inservice Testing/Calibration, and Administrative Controls and Engineering support.

Plant management's awareness of, involvement in, and support of safe plant operations were also evaluated. Emphasis was placed on interviews of personnel at all levels, observations, and system

N ll 1

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 39 walkdowns. The inspectors also reviewed plant deviation reports, Licensee Event Reports for the current SALP period, and evaluated the effectiveness of the St. Lucie's root cause identification, short-term and programmatic corrective actions, and repetitive failure trending and related corrective actions.

Within the areas inspected, several strengths were noted, however, some weaknesses and enforcement items were also identified. Corrective actions to address enforcement items have been completed.

4.6.3 Maintenance Team Inspection In October and November 1989, NRC conducted a Maintenance Team Inspection at the St. Lucie Plant. The results of the inspection are documented in NRC Inspection Report 89-24 dated January 23, 1990..This special announced team inspection consisted of an in-depth inspection of the maintenance program and its implementation.

The overall results determined the maintenance program to be good with good implementation.

Significant strengths were identified in the report, The NRC recognized that plant management is involved in the maintenance program and appeared to be aggressive in improving the program.

Commitment of engineering for a new maintenance facility, the existing training facility, other planned program improvements, and receptiveness to team findings indicated a positive attitude toward program improvements. Generally, the maintenance staff and supervision were knowledgeable and qualified. Work observed was accomplished in an organized professional manner with emphasis on following procedures. In particular, the I&C and Electrical groups were found to be strong.

Two violations were identified regarding control of class IE panel components and discrepant maintenance records. Corrective actions to address identified enforcement items have been completed.

4.6.4 Equipment Environmental Qualification Inspection In February and March 1989, the NRC conducted an inspection of the 10 CFR 50.49 requirements for environmental qualification (EQ) of electrical equipment at St. Lucie Plant, The inspection included a review of the FPL implementation program to meet the requirements of 10 CFR 50.49, walkdown inspections of EQ equipment inside containment, review of EQ maintenance activities, review of EQ design changes and FPL actions in response to NRC initiatives (NRC Notices 86-71 and 88-89), and follow-up on FPL actions in response to license conditions.

The results of the inspection were documented in NRC Inspection Report 89-07 dated April 6, 1989. No violations or deviations were identified. The results of this inspection supported the NRCs previous assessment in March 1986, that FPL had implemented an adequate program and that the program continues to be adequate. The walkdowns of EQ equipment resulted in no open

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 40 or unresolved items. EQ documentation files were considered well organized and complete, requiring minimal additional information to support the analysis provided in the file. EQ maintenance program procedural compliance was good, and the special maintenance requirements described in the EQ document packages were confirmed to have been incorporated into the procedures for those sample items reviewed.

The inspection team noted that the tracking mechanism used by the plant to ensure scheduling and completion of EQ maintenance tasks appeared to be fragmented with separate lists for calibration, surveillance, preventive maintenance (PM), and replacement. There was no overall program to ensure that maintenance activities were being accomplished. This was considered a weakness in FPL's EQ maintenance program. FPL acknowledged the concern and committed to implement a formalized maintenance tracking system as outlined in the inspection report details. The inspection team considered these actions to be adequate to resolve concerns.

This inspection supports FPL's position that St. Lucie equipment under the EQ Program was performance tested and maintained in accordance with the plant design bases.

4.6.5 Conclusion from NRC Inspection Efforts These inspections evaluated the St. Lucie processes for operating and maintaining the plant as designed. Plant configuration, SSC maintenance, component testing and plant operation were evaluated by these teams and found to be acceptable. Deficiencies noted by the inspectors were entered into the corrective action program for resolution as necessary. Follow-up inspections verified the corrective action process. Therefore, these inspections support the rationale that the processes in place willassure that design basis information is maintained in maintenance, operating and testing procedures 4.7 FPL Audits, Self-Assessments and Findings The effectiveness of St. Lucie initiatives, projects, programs, and processes have been audited by FPL QA. Provided below and in Appendix B, Table B 4.7.1 are details of selected QA audits.

The overall review of these audits supports the conclusion, with reasonable assurance, that St.

Lucie design bases requirements have been maintained and properly translated into operating, maintenance, and test procedures, and that discrepancies found are corrected in a timely manner.

4.7.1 Audit Findings and Strengths Since the initial startup of the units, numerous formal audits have been performed by the Quality Assurance organization in areas related to the control of design bases, design, configuration, testing, and plant procedures at St. Lucie. Design program audits typically reviewed engineering processes, implementing procedures, and the implementation of design changes and associated records (i.e. modification packages). QA audits consistently found that Quality Assurance Program elements (including design requirements) were adequately addressed by procedures, and that the implementation of those procedures was effective, The programs and procedures for

c l hg' St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 41 translating design requirements into procedures and instructions were consistently determined to be adequate. Audit Findings were resolved through the corrective action process, and resulted in improvements of Engineering, Operations and Testing processes and procedures.

One trend noted by QA was repeated instances of less than adequate translation of design bases into procedures and drawings, In 1996, QA Audit QSL-PCM-96-11 identified that there were insufficient procedure requirements to ensure that procedures affected by design changes or safety evaluation requirements were identified and updated prior to implementation. The three Findings issued in Audit QSL-PCM-96-11 resulted in improvements to the modification package implementation processes to include a Procedures Writing Group review of the PC/M prior to implementation. Other relevant examples of QA audits which prompted plant and process improvements are highlighted in Appendix B, Table B 4.7.1, "QA Audit Findings," and are summarized below.

~ QAS-JPN-91-8, "Juno Nuclear Engineering - Production Engineering Group (PEG) for St. Lucie."

~ QAS-JPN-92-3, "Nuclear Engineering."

~ QAS-JPN-93-3, "Nuclear Engineering - St. Lucie Design Control."

~ QAS-JPN-95-1, "Nuclear Engineering Audit."

QSL-PM-96-18, "Performance Monitoring Audit."

Other QA audits and self-assessments listed in Appendix B, Table B 4.7.1 resulted in findings which directly related to the proper translation of design bases requirements into operating, maintenance and testing procedures. Resolution of findings from the following audits listed in Appendix B, Table B 4.7.1 promoted significant improvement in the St. Lucie design basis maintenance and procedure consistency processes:

~ QSL-PM-96-06, "Performance Monitoring Audit." Discrepancies between the FSAR and St. Lucie implementing procedures were identified and procedures did not adequately address design requirements.

~ QSL-PM-96-08, "Performance Monitoring Audit." A discrepancy between the FSAR and St. Lucie implementing procedures was identified.

Corrective action for these findings also contributed to the process change to add the Procedures Writing Group to the PC/M process for review of affected procedures before the PC/M is approved for implementation.

4.7.2 1996 Plant Self-Assessment In October 1996, St. Lucie completed a five and one half month self-assessment of site performance during the two year period from January 1994 to May 1996. On October 18, 1996, the Self-Assessment Team published a comprehensive report which documented their evaluation of the performance of operations, engineering, maintenance, and plant support functional areas, as well as management policies and the corrective action program. The evaluation of the

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 42 engineering functional area included engineering design control and configuration management.

The team concluded that the overall quality of engineering work was superior. However, plant drawings, procedures and licensing documents were not always updated to reflect changes to the plant. In addition, guidance for updating operating procedures, as-built documents, and the FSAR, to reflect changes due to plant modifications, was unclear. The issues were addressed in the summer of 1996 with the implementation of new guidelines for implementing plant modifications.

The complete history of QA audits and self-assessments has resulted in continual process improvements in design and configuration control, and in plant procedure development and compliance with the plant design bases.

5.0 Conclusion There is adequate rationale for concluding with reasonable assurance that the operations, maintenance, and testing procedures at St. Lucie are consistent with the design and that discrepancies between design documents, the FSAR, procedures and the plant are not significant with respect to safety based upon the facts that:

~ The verification at initial licensing of the plant, FSAR and the design are consistent.

~ The processes in place assure that consistency is maintained.

~ Various improvement projects and programs throughout the history of the plant confirm the acceptability of plant procedures.

~ The various audits, inspections, and self-assessments have verified consistency of design with the plant and its translation into the procedures.

~ Processes have been and are in place for identification, evaluation and correction of discrepancies between the design and the plant procedures.

St. Lucio Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 43 Ic] "Rationale for concluding that system, structure, and component configuration and performance are consistent with the design bases."

An outline of the organization of the response to this request follows:

1.0 Introduction 2.0 St. Lucie Programs and Initiatives 2.1 Design Basis Program 2.2 Containment Penetration Review 2.3 LOCA Containment Re-analyses 2.4 Instrument Setpoint Verification 2.5 Unit 1 Steam Generator Replacement Project (SGRP) 2.6 Simulator Validation Project 2.7 Operating Experience Feedback (OEF) Program 2.8 Surveillance Testing 2.9 Maintenance Rule Program 2.10 Plant Procedure Improvement Program 2.11 Annunciator Summary Review 3.0 Walkdowns, Inspections, Audits and Reviews 3.1 FPL Plant Walkdowns 3.2 FPL Seismic Walkdowns 3.3 FPL Unit 1 Steam Generator Replacement Walkdowns 3.4 NRC Electrical Distribution Functional Inspection (EDSFI) 3.5 NRC Service Water System Operational Performance Inspection 3.6 NRC Inspection of Plant Changes and Modifications 3.7 NRC Integrated Inspections 3.8 NRC Equipment Environmental Qualification Program Inspection 3.9 NRC Inspection of the Motor Operated Valve Program 3.10 NRC Check Valve Inspection 3.11 NRC Architect/Engineer Team Inspection 3.12 FPL Audits and Reviews 4.0 Conclusion

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 44 1.0 Introduction The rationale for concluding with reasonable assurance that the St. Lucie Plant system, structure, and component (SSC) configuration is consistent with the design bases is based on the following:

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

2. Startup testing verified plant performance conformed with the design bases for plant operation.

3. Plant changes have been under procedural control to assure consistency with the design bases since initial operation.

4. Process improvement initiatives have resulted in improved effectiveness of the processes used to maintain the plant configuration and performance consistent with the design bases.

5. Routine audits, inspections and vertical slice reviews confirm that plant performance and configuration is consistent with the design bases.

St. Lucie Units 1 and 2 were licensed in 1976 and 1983, respectively. The plant configuration was verified to be consistent with design information and the FSAR prior to initial operation. The original design process produced design documents such as calculations, drawings, specifications, evaluations, and analyses necessary to support initial construction, testing, operation, and licensing of the plant. The design documents produced by this process were used as the bases for system startup and acceptance testing. The design and construction process included various quality assurance audits, quality control inspections, and documentation requirements to assure consistency between the as-built systems, structures, and components, and the design.

The FSAR was produced as part of the plant design process and submitted to the NRC for review and approval. The FSAR includes sufficient descriptions of the plant's design, configuration and performance requirements to document compliance with NRC regulations. NRC review of the original FSAR resulted in amendments prior to plant operation, The NRC Safety Evaluation Report documents this review and the staff's findings that the plant's design, configuration, and performance parameters were acceptable. Technical Specifications were developed (in conjunction with the NRC review) to identify configuration and functional requirements, controlling parameters, and surveillance and testing requirements that are significant to the safe operation of the plant. These licensing, design, and as-built configuration reviews, conducted in conjunction with the initial startup testing of the plant, confirmed that the system, structure, and component configuration and performance were consistent with the design bases prior to initial operation.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 45 Startup testing verified plant performance conformed with the design bases for plant operation.

The initial startup testing program was developed based on design basis requirements identified by the NSSS supplier and the A/E. Abstracts for the startup test procedures were provided to the NRC for review and acceptance, The successful completion of the initial startup test program verified that the plant's licensed configuration and performance were in compliance with its design bases. A startup test report was submitted to the NRC prior to commercial operation. This report formed the basis for subsequent plant operation and inservice testing.

Since startup, plant changes have been under procedural control to assure consistency with the design bases. As discussed in the response to Request [a], the evolutionary and current engineering design and configuration management processes have provided the necessary controls to ensure that the St. Lucie SSC configuration and performance have been and remain consistent with the design bases. In addition, FPL has sufficient documentation (calculations and/or pre-operational, startup, and surveillance test data) to conclude that the current plant configuration is consistent with its design bases.

The FPL engineering design and configuration control processes were developed and maintained to ensure that the plant design, as-built condition, and operation remain in conformance with the original plant license, design bases, and subsequent licensing commitments. These processes are used to control the plant design basis, design, design changes, physical configuration, and operations, maintenance, testing, installation, procurement and training requirements and documentation.

2.0 St. Lucie Programs and Initiatives St Lucie programs and initiatives support the rationale that SSC configuration and performance are consistent with the design bases. Together they have helped identify process improvements, correct procedural/plant configuration inconsistencies, improve content and accessibility of design bases information, and improve plant performance. These programs and initiatives, which support the rationale that systems, structures and component configuration and performance are consistent with the design bases, are discussed in the following sections, 2.1 Design Basis Program The Design Basis Program involved the development of a set of unit-specific system-level documents, referred to as design basis documents (DBD), that contain a roadmap to the reference regulations, codes, standards, calculations, analyses, specifications, etc., that form the basis for system design, testing and operation. Each DBD was reviewed against plant drawings in an effort to ensure consistency with the design bases, and the plant configuration and operating procedures.

The DBDs explain system design and provide a definition of bases for the design, component design constraints, and design features. The results of the Design Basis Program support the conclusion that SSC configuration and performance are consistent with the plant's design basis.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 46 2.2 Containment Penetration Review In 1991, an evaluation, "Technical Assessment of Containment Penetration Boundaries," was issued for each unit to address plant events related to the control of containment isolation boundaries. The objectives of these evaluations were to:

1) perform a technical assessment of each mechanical containment penetration to determine those components whose integrity or operational position were vital to containment integrity; and

2) provide sketches identifying containment boundary components to assist plant personnel involved in operations and maintenance activities.

The evaluations documented the containment penetration boundary, including components that are vital to containment integrity.

In preparing the evaluations, each penetration configuration was evaluated to:

1) confirm that the intent of the 10 CFR 50 Appendix A, General Design Criteria (GDC) requiring a double barrier is satisfied for each line penetrating the containment; and

2) identify those components that form the containment penetration boundary, including those components that can be manipulated (e.g., valves opened, caps removed, or piping blind flanges removed) such that the double barrier criterion could be violated.

Appropriate design and construction documentation was consulted and, where possible, the system components were walked-down to generate and confirm the containment penetration sketches.

The Unit 1 evaluation concluded that the isolation system meets the intent of the 10 CFR 50 Appendix A, General Design Criteria (GDC) in that a double barrier is provided in each line penetrating the containment. However, certain containment penetration valving did not conform in detail to the requirements of GDC 54 through GDC 57, since the General Design Criteria were published after the Construction Permit was issued, and hence, were not available as a guide.

The Unit 2 evaluation concluded that the isolation system meets the requirements of the 10 CFR

. 50 Appendix A, General Design Criteria (exceptions taken to GDC provisions are discussed in Section 6.2.4.3 of the St Lucie Unit 2 UFSAR) in that a double barrier is provided in each line penetrating the containment.

~

2.3

~ LOCA Containment Re-analysis In 1993, the LOCA containment analyses of record were updated. Prior to the update, the

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containment section of the FSARs included much of the original design basis sensitivity studies.

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St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 47 Plant changes, (e.g., power uprate), had been addressed, however, this information was typically updated with the most limiting case of record, and as a result, portions of the FSAR had become unclear. The re-analyses and associated FSAR updates, incorporated changes into one evaluation for each unit. It utilized operating experience where necessary to support inputs including the final product of peak and transient containment pressure and temperature, and input for the Design Basis Program.

The results of the re-analyses showed containment pressure to be within both the peak containment design pressure and the Technical Specification integrated leak rate test pressure.

The peak containment temperature was also well within the peak containment vessel temperature limit. The long term temperature response was also shown to be within the existing equipment environmental qualification (EQ) envelope for Unit 2. Although the long term temperature response for Unit 1 was shown to slightly exceed the existing equipment EQ envelope for a limited time interval, when the equipment test curves were reviewed, they bounded the new analyses, and the EQ curve was revised.

Additionally, the re-analyses evaluated the CCW temperature response using a time dependent containment heat load. The CCW and ICW loops were conservatively modeled without credit for the time lags associated with loop transit times and piping volumes. In this manner, a realistic approach to peak CCW temperature calculations was conducted, thereby providing an accurate baseline of predicted CCW temperature for any future analyses, and valuable input for plant operation.

2.4 Instrument Setpoint Verification Between 1991 and 1993, FPL performed an instrument setpoint review for the St. Lucie Plant to verify and document current setpoints. Approximately 250 safety-related and important-to-safety instrument loops were evaluated with approximately 150 loops requiring documentation or recalculations. Plant operating and maintenance procedures were compared to the current field conditions and design documents to verify consistency. There were no design basis inconsistences found which were of a magnitude which required NRC notification. This program provides a basis to conclude that the current plant setpoints are correctly translated into plant procedures and that the design basis is available.

The original St. Lucie instrument setpoints were furnished by the A/E, the NSSS designer, or other equipment vendors, and have been supplemented through the years by engineering packages, which were typically produced by the A/E. Prior to the instrument setpoint verification effort, control of instrument setpoint changes was governed by a separate process that involved plant and engineering procedures. In addition, it was an accepted practice at the time that some of the setpoints could be changed by the plant based on A/E or vendor input with limited documentation. Historically, the setpoint bases have been maintained in various documents including drawings, vendor manuals, and maintenance procedures, and the calibration/scaling values (typically calculated by maintenance) were maintained in the calibration procedure.

, I St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 48 The setpoint verification effort (in addition to documenting or recalculating instrument setpoints) changed the old process by incorporating instrument setpoint changes into the configuration management process. The process change converted the setpoint list into an engineering drawing and placed instrument setpoints under engineering control in the configuration management process. Bases documents such as calculations were updated as appropriate to reflect the instrument setpoints described in the design drawings.

2.5 Unit 1 Steam Generator Replacement Project (SGRP) 2.5.1 Evaluation of Accident Analysis The accident analyses presented in Chapter 15 of the St. Lucie 1 FSAR were evaluated to determine the effects of use of the replacement steam generators (RSG). The objective of each evaluation was to demonstrate that the St. Lucie 1 plant response with the RSGs would meet all NRC approved FSAR acceptance criteria.

The specific transients considered in the Stand-Alone Safety Evaluation (SASE) were grouped according to the categories described in NRC Regulatory Guide 1.70, with a cross reference to the St, Lucie 1 FSAR. The categories are:

e 1. Increases in heat removal

2. Decreases in secondary heat removal

3. Decreases in reactor coolant system flow

4. Reactivity and power distribution anomalies.

5. Increases in reactor coolant inventory

6. Decreases in reactor coolant inventory

7. Radioactivity releases from subsystems The FSAR was reviewed to determine the effects of using the RSGs. The safety evaluation of the FSAR Chapter 15 accident analyses, the main steam line break and loss of coolant accident analyses for reactor building pressure response in FSAR Chapter 6, and the overpressure protection and natural circulation analyses in FSAR Chapter 5 are also discussed in SASE.

2.5.2 Evaluation of Reactor Building Response The RSGs can affect the mass and energy releases for primary and secondary system ruptures.

Consequently, the effects of use of the RSGs on the St. Lucie 1 reactor building pressure calculations in Chapter 6 of the FSAR were assessed.

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2.5.3

~ ~ Evaluation of Structural Response The loadings resulting from a LOCA, Main Steam Line Break and seismic events were evaluated for the RSGs and reactor coolant system. The objective of these evaluations was to verify that

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St. Lucic Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 49 the existing components, supports, and restraints were adequate to accommodate the RSGs under normal and accident loading conditions. The evaluations addressed all primary components.

2.5.4 Evaluation of Other RCS Interfaces Generally, the use of the RSGs does not affect the balance of the reactor coolant system or its supporting analyses. However, certain interfaces with other plant features were evaluated with respect to use of the RSGs. Items considered include:

1. Effects of RSG hydraulics on core lift

2. Nozzle dams

3. Overpressure protection analyses in Appendix 5A of the FSAR

4. Low temperature overpressure protection (LTOP) analyses in Appendix 5B of the FSAR

5. Natural circulation analyses in Appendix 5C of the FSAR

6. Internal missile evaluation in Section 3.5 of the FSAR

7. Loose parts

8. Feedwater control 2.5.5

~ ~ Evaluation of Plant Documentation Comprehensive reviews of the St. Lucie 1 Technical Specifications, FSAR, plant emergency and off-normal operating procedures, and the existing stress analysis were performed to identify changes that may be required due to use of the RSGs. The technical specification review included limiting conditions for operation, surveillance requirements, and bases. The FSAR review included a page-by-page review of the FSAR text and tables.

2.5.6 SGRP Project Specific Design Basis Documents The SGRP developed specific design basis documents for the project special processes and other project specific tasks.

Rigging and handling Cutting and beveling Steam generator replacement welding Temporary supports/restraints and structures Blowdown system interface Replacement steam generator insulation Steam generator manway platforms RCS decontamination These documents were used during the development of the various project engineering packages, The applicable St. Lucie DBDs willbe utilized, along with other appropriate bases documents, during the 1997 reconciliation phase of the SGRP design documents prior to implementation.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 50 2.5.7 Conclusion The evaluation of the existing plant design and licensing bases, steam generator physical interfaces, thermal-hydraulic aspects, structural supports, safety analyses, technical specifications, and off-normal and emergency procedures showed the RSG design to preserve the existing plant design and licensing bases. Therefore, use of the RSGs at St. Lucie 1 meets all relevant criteria and guidance, the existing licensing and design bases remain valid, the Technical Specifications and their bases remain valid, and no unreviewed safety questions exist.

2.6 Simulator Validation Project St. Lucie has a reference plant simulator. It was developed as an operator training tool and is based on extensive reviews of the design, design bases, and, operational data and procedures, to ensure consistency between the actual plant and the response of the simulator. In instances when the simulator predicts an unexplained plant response, the feedback process in the training programs provides for resolution of differences.

Another use of the simulator is to validate that proposed design changes to the plant are safe and effective. By first simulating major changes, St. Lucie has been able to identify and correct design differences prior to implementation into the plant. The simulator is also used to improve operator understanding of the plant response to unexpected events, such that corrective actions are identified, and contributors, or responses that may be preventative, are also identified.

The Training Department assists the Operations Department in validation of procedures by using draft procedures during the simulator portion of licensed operator continuing training. The draft procedures are provided to training by the Operations Department and the feedback from the crews are forwarded to the procedure development group.

2.7 Operating Experience Feedback (OEF) Program Industry events and reports'are reviewed and processed based on applicability to St. Lucie in accordance with the Operating Experience Feedback (OEF) Program. Operating experience events and reports are reviewed and screened for applicability to St. Lucie by the OEF Coordinator.

Operating experience feedback which merit analysis, specific actions, or a documented plant review or response, are processed as Condition Reports. The Condition Report process ensures thorough analysis of the event and preparation of a report such that relevant issues are addressed including update to the plant documentation ifrequired.

Operational experience feedback is used to improve plant processes, and where process enhancements improve plant safety and/or operation, procedures are revised. The OEF program endeavors to minimize the potential for an initiator of an event not being recognized in the change process of a procedure.

'I M St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-2S Enclosure Pago 51 2.8 Surveillance Testing Surveillance testing is used to verify the operability of SSCs by ensuring their performance remains within the specified design bases, The identification of degraded equipment prior to failure increases equipment availability and assures that components required to operate during an accident are operable. Surveillance requirements are obtained from the FSAR, Technical Specifications, ASME Boiler and Pressure Vessel Code,Section XI Pump and Valve program, and other criteria specified as necessary. Programs are in place to assure failed or missed surveillances are appropriately evaluated for reportability and necessary corrective actions are implemented.

2.8.1 Technical Specification Surveillance Testing Technical Specification surveillance testing provides assurance that SSCs operate to perform their intended function. Surveillance testing is accomplished and tracked by operating and administrative procedures. Components which fail the surveillance tests are entered into the condition reporting process for root cause determination and resolution. This process identifies degraded equipment and allows timely repairs.

~ ~

2.8.2

~ ~ Inservice Testing Program

~ ~

The inservice testing is used to verify the operational readiness of pumps and valves which have a specific accident mitigating or safe shutdown function.

The general requirements of ASME Paragraphs IWV-1100 and IWP-1100 for Class 1, 2 and 3 pumps and valves apply to St. Lucie. These requirements establish the inputs used in the FSAR safety analyses. The following describe SSC functions, and were reviewed for the Inservice Testing Program scope:

Updated Final Safety Analysis Report Technical Specifications System descriptions for Training Special analyses Commitment correspondence Plant procedures 10 CFR 50, Appendix J leakrate test program The ASME Inservice Testing Programs for St. Lucie Units 1 and 2 are for the second 10-year interval and will be in effect through the end of the second 120-month (10-year) interval unless revised and reissued for reasons other than the routine update required at the start of the third interval, The NRC has reviewed and approved these test programs.

The ASME Inservice Testing Program tests the specific as-built configuration values (i.e.,

individual or integrated component characteristics/parameters for each pump and valve) used in

St. Lucio Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 52 the FSAR safety analyses. In this way, the specific system and component capabilities used in the safety analyses are better assured under all plant conditions.

2.9 Maintenance Rule Program The System and Component Engineers implement the 10 CFR 50.65 maintenance rule program.

System engineers monitor the performance of systems and system components against set criteria.

Maintenance activities to correct degraded equipment are monitored. Maintenance preventable functional failures (MPFF), a key indicator, are tracked and trended and entered into the condition reporting process for root cause evaluation and resolution. System engineers perform the function failure evaluation in accordance with administrative procedures. The program administrator has the responsibility to track, monitor, and identify repetitive MPFFs and to return the system, structure, or component to its prescribed acceptable limits.

2.10 Plant Procedure Improvement Program The St. Lucie Plant Procedure Improvement Program is focused on regulatory performance improvement. The scope of the project is to identify and improve operation and maintenance activities by upgraded task instructions, Included in the upgrade process is review of the design basis documents to ensure regulatory commitment compliance.

The Procedure Improvement Program includes operations, maintenance, and administrative procedures. The initial step included centralizing procedure writers from Operations, Electrical Maintenance, Instrument and Control Maintenance, Mechanical Maintenance, and Administrative Departments under a procedure development supervisor. This new group provides a single point of contact for procedural issues. The goal of this new group was to upgrade procedures as necessary and to prepare new procedures for tasks without procedures. The basic format for procedures willbe standardized and use industry-proven language techniques to give the end-user a technically correct, first time implementable procedure.

The improved process includes the following fundamental attributes:

~ Define the scope of a technical subcommittee

~ Review procedures for critical characteristics and attributes

~ Include industry standard methods for procedure validation

~ Validate by end-user to ensure the procedure is sound and implementable

~ Establish a site procedure writer's guide which incorporates INPO and industry approved writing standards Another fundamental attribute of the improved process is to review, as appropriate, design basis documents that govern the activities and control the tasks described in each procedure. These basis documents include, but are not limited to, St. Lucie Technical Specifications, FSAR, Off-site Dose Calculation Manual, NRC correspondence, ABB-CE bulletins, calculations, analyses, evaluations, specifications and plant drawings.

I St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 53 2.11 Annunciator Summary Review A plant annunciator summary review is currently in progress. Annunciators provide an early indication/announcement of possible degrading plant conditions, and as such they are an important aid to operators in identifying appropriate procedural actions. This program willreview selected annunciators for the control rooms, waste management panels, emergency diesel generators, steam generator blowdown building, boric acid concentrators, waste concentrators, water treatment plant, oxygen analyzer, and the liquid waste panels. This program willinclude review of selected logic configurations, annunciator setpoints, procedural references and corresponding operator actions found in the annunciator summary procedure for each unit, against the design documents, as appropriate. This project is approximately 10% complete with over 200 annunciator setpoints, configurations and procedural responses reviewed, and has not identified any inconsistencies which would be considered safety significant. This program continues to support the conclusion that the SSC configuration and performance is consistent with the plant's design basis.

3.0 Walkdowns, Inspections, Audits and Reviews There have been numerous plant walkdowns, NRC Inspections, and FPL Audits and Vertical Slice Reviews, which have verified consistency between the SSCs and the design bases.

3.1 FPL Plant Walkdowns As one means of ensuring that the plant configuration is consistent with the design bases, plant walkdowns are performed as part of the System Engineer Program. System Engineers are cognizant of their system's operating status, including scheduled and unscheduled maintenance, periodic testing and results, system arrangement and operational configuration, system performance margins, and performance trending.

3.1.1 System Engineer Walkdowns AP 0005750, "Duties and Responsibilities of the System Engineer," is the governing document which provides the basis for the system engineer program. The system engineer is the "owner" of the plant system and remains cognizant of the systems design bases, its performance criteria, system operation, and maintenance requirements. The system engineer reviews proposed plant modifications to the system. Material condition is monitored by system walkdowns.

Engineers perform system walkdowns in accordance with Guideline SCEG-003, "Guideline for the Condition Survey of Structures and Supports by Plant Personnel." The results of these walkdowns are used to demonstrate compliance with the maintenance rule. Procedural and drawing inconsistencies identified by the system engineers are entered into the CR process for resolution.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 54 3.1.2 QA Walkdowns QA safety system walkdowns provide another means of ensuring that the plant configuration is consistent with the design bases. Since January 1996, walkdowns are conducted on a quarterly basis as part of the QA Technical Reviews and Assessments (TRA) group functions. The TRA group was formed to encompass the former ISEG functions, which focuses on issues affecting nuclear safety and plant reliability. Prior to 1996, QA performed vertical slice audits of safety systems, which also included system walkdowns. As noted in Appendix B, Table C 3.12 the walkdowns have helped verify plant configuration to plant drawings and procedures. During some of these audits, minor inconsistencies between the plant and design documents were identified and entered into the corrective action program for resolution.

3.2 FPL Seismic Walkdowns 3.2.1 IE Bulletin 79-14 IE Bulletin 79-14, "Seismic Analysis for As-Built Safety-Related Piping Systems," addressed an NRC concern that the seismic analysis of safety related piping, coupled with pressure, thermal, operating weight, and other applicable system operating parameters, be verified in accordance with actual as-built drawings for its impact on the stress analysis of record. At St. Lucie Units 1 and 2, the verification process consisted of a program by which the as-built configurations of Seismic Category I safety-related piping systems were documented, evaluated, and modified as necessary, by a team of experienced engineers and designers under the FPL QA/QC program.

This program consisted of walkdowns of designated Seismic Category I safety related piping systems and documentation of the as-built configuration including piping size, components, and geometry; location, orientation, and type of valves; location and configuration of pipe supports; and floor and wall penetration details. These details were then evaluated by the team to assess if document revisions and/or plant modifications were required and, if necessary, acceptable under the construction code of record.

Commitments made by FPL with regards to NRC Bulletin 79-14, which included walkdowns, package preparation, evaluation, and modifications, were completed for St. Lucie Units 1 and 2.

3.2.2 Generic Letter 87-02 Generic Letter 87-02, "Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactor," (USI A-46), required a review of St. Lucie Unit 1 equipment required to achieve and maintain hot shutdown of the plant for a period of eight hours. FPL selected safety-related and non-safety-related equipment for their review. This equipment was evaluated for seismic adequacy by FPL engineers and a seismic review team consisting of engineering experts in the area of seismic adequacy of equipment and equipment performance during earthquakes. The four basic requirements for the GL 87-02/USI A-46 review of St. Lucie Unit 1, which were

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 55 addressed in the walkdown are; the equipment seismic capacity being greater than demand, the construction adequacy of the equipment, anchorage adequacy and no seismic spatial interaction.

The walkdowns resulted in the identification of three equipment items, and two other items which were acceptable but required maintenance on the anchorage. Modifications were made to all equipment identified with deficiencies, and a report was issued to the NRC in 1992. By letter dated February 9, 1995, the NRC issued a Safety Evaluation Report (SER) addressing the St.

Lucie and Turkey Point response to 87-02 with several outstanding issues. An audit was performed at Turkey Point in December of 1995, and the NRC issued a supplemental SER on October 22, 1996, addressing outstanding issues found during the audit. In this SER, open items for St. Lucie Unit 1 were closed.

3.3 FPL Unit 1 Steam Generator Replacement Walkdowns A 10 CFR 50.59 safety evaluation an~ ",t of design packages to be used during the 1997 steam generator replacement outage have b veloped. These documents are now undergoing a design review and verification process prior to implementation. Plant walkdowns were conducted in the course of doing this design work, and the plant FSAR, design analysis and other applicable licensing basis documents were reviewed.

Walkdowns were conducted inside and outside containment in the areas affected by the SGRP.

The specific areas of interest were temporary structures which interface with permanent plant structures and where interferences will need removal to accommodate movement of the new steam generators. During the walkdowns, minor pipe and HVAC support discrepancies between the design drawings and the as-found condition, were documented, evaluated and dispositioned.

In all cases physical changes were not necessary and the appropriate drawings were modified. An input error in the FSAR accident analysis was documented, evaluated and corrected by the source fuel vendor, and determined not to affect the conclusion of the accident analyses.

3.4 NRC Electrical Distribution System Functional Inspection In February and March 1991, the NRC conducted a special announced team inspection in the areas of design of electrical systems and related engineering and maintenance activities. This inspection specifically focused on the Electrical Distribution System (EDS) as-built configuration conformance to design bases requirements and design output documents. The inspection included a review of design, calibration, maintenance, and the "as-built" configuration of the electrical distribution system including mechanical systems and equipment associated with the EDS.

3.4.1 FPL Preparation for the EDSFI St. Lucie Engineers and Plant Staff expended over 13,000 man-hours between 1990 and early 1991 in preparation for the inspection and conducted two self audits. Key electrical design and configuration aspects of St. Lucie Units 1 and 2 were reviewed and updated as required. Some of the areas were:

~ ~

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 56

~ Electrical system short circuit and voltage drop analyses

~ Battery and battery charger sizing

~ DC system minimum and maximum voltage

~ Breaker coordination

~ Fuse sizing

~ Cable sizing

~ Protective Relay settings

~ Emergency Diesel System - Electrical design 8c accident loading

~ Emergency Diesel System - Mechanical design, fuel 8c cooling system

~ As-installed configuration versus design drawings Upon completion of this effort, several calculations were updated or revised. In addition, a number of inconsistencies were identified in the FSAR and plant configuration drawings. These inconsistencies were scheduled for correction. During the FPL review, no operability problems were identified with the St. Lucie electrical distribution system.

3.4.2 NRC Inspection The NRC conducted an announced team inspection during the first quarter of 1991. During this extensive technical audit, the following areas were reviewed by the team:

~ Transmission System - The team reviewed the characteristics of the 240kV electrical transmission system. Maximum and minimum range of voltage and frequency was reviewed.

~ Medium Voltage (4.16KV) Class 1E Electrical System - a review of the 4.16kV and emergency diesel design was conducted. Many of the calculations and analyses for equipment loading, short circuit, and voltage regulation were reviewed. Also included was a review of the diesels'apability to start and accelerate the assigned safety loads within the required time sequence. The team concluded the maximum calculated fault current levels and voltage regulation for the system was demonstrated to be acceptable in all design conditions.

~ Class 1E 480 VAC System - short circuit and voltage drop calculations were reviewed for the 480Vac electrical system. Short circuit ratings were found to be within the equipment rating. Cables were adequately sized and breaker coordination was acceptable.

~ Class lE 120 VAC System - the Class 1E 120 Vac design was reviewed. Several short circuit and voltage drop calculations were performed to demonstrate that cable sizing was acceptable for the equipment.

~ Class 1E 125 VDC System - reviewed design documentation, calculations, and evaluations which demonstrated sizing and loading for the 125 Vdc system and

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 57 equipment. This documentation included short circuit and low voltage calculations, MOV design studies, high voltage analysis, and coordination studies. The documentation appropriately verified dc equipment was adequately sized. Minor documentation errors were identified by the team. FPL investigated the errors and agreed to correct the documents in the forthcoming revisions.

~ DC Motor Operated Valves - voltage drop calculations for dc motor operated valves were found not to meet present day vendor recommendations. Additional information supplied by the vendor confirmed that acceptable torque was provided to operate valves under plant stated conditions. The team concluded that no operability concern existed in this area.

~ Maintenance, Testing, Calibration, and Configuration control - various walkdown inspections were conducted to verify the "as installed" configuration of the electrical system and determine compliance with the design drawings and documents, Completed calibration and surveillance procedures were also reviewed to verify that system functions were tested in accordance with design specifications. Although the team identified some minor issues, it was determined that a comprehensive program for the maintenance, testing and calibration of the electrical equipment was in place at St. Lucie. Overall, the team determined the results of the walkdown inspections were very satisfactory.

The results of this inspection were documented in NRC Inspection Report 91-03 dated May 17, 1991, and no violations, deviations or operability concerns were noted. The NRC concluded that the Electrical Distribution System (EDS) at St. Lucie was capable of performing its intended function under normal and accident conditions, and that adequate controls are in place to maintain the EDS in an operable configuration. The engineering technical support and design groups supporting the EDS were also found to be adequate. Their performance demonstrated involvement in problem identification and resolution as well as routine activities in maintenance, testing, operations, and procurement. Plant modifications to the EDS were performed in accordance with the approved design control processes.

Minor discrepancies were noted in the areas of the startup transformer, the non-safety cooling fan in the emergency diesel generator rooms, fuse control and relay setting drawings, the relay setting calibration procedure and preventive maintenance to detect failed molded case circuit breakers.

These discrepancies were resolved under the St. Lucie Corrective Action Program.

The results of this inspection demonstrated that the EDS at St. Lucie willperform its intended function under normal and accident conditions and that FPL had adequate controls in place to maintain the EDS in an operable configuration.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 58 3.5 NRC Service%'ater System Operational Performance Inspection In September and October 1991, the NRC conducted a special announced pilot team inspection in the area of the service water operational performance. This inspection specifically focused on the service water (i.e intake cooling water at St. Lucie) system as-built configuration and demonstrated operational performance regarding the applicable design bases requirements. The inspection included a mechanical design review; detailed system walkdowns; review of system operation, maintenance, and surveillance; and assessment of quality assurance and corrective actions related to the Intake Cooling Water (ICW) system. The team also assessed FPL's implementation of actions required by Generic Letter 89-13, "Service Water System Problems Affecting Safety-Related Equipment," as well as system unavailability to gain additional insights for probabilistic risk assessment applications.

The results of the inspection were documented in NRC Inspection Report 91-201. The NRC inspection team concluded that the Service Water System as-built configuration would be capable of performing its intended safety function under design bases conditions. Discrepancies were noted in the areas of testing the standby "C" pump, ICW temperature control valve stroke time testing, IST of ICW manual valves, FSAR inconsistencies and training material inconsistencies.

Overall, this NRC SWSOPI found no discrepancies in the plant design bases, however, the review contributed to improvements in the St. Lucie processes for maintaining its plant configuration, performance, and testing procedures consistent with its design bases.

3.6 NRC Inspection of Plant Changes and Modifications In November 1993, the NRC conducted a routine announced inspection of the area of design changes, plant modifications, and engineering and technical support activities. This inspection specifically focused on the adequacy of the Plant Change/Modification (PC/M) process to ensure that changes to the plant design and/or as-built configuration were implemented such that the modified configuration and system/component performance were in compliance with the plant design bases requirements.

The results were documented in NRC Inspection Report 93-25, dated December 1, 1993. In the areas inspected, no violations or deviations were identified. FPL had demonstrated the use of an adequate prioritization process for identifying and implementing plant modification. The overall quality and technical content of the PC/M packages reviewed were adequate and sufficiently documented to verify closure. The 10 CFR 50.59 safety evaluations provided sufficient discussion and justification to comply with regulatory requirements. FPL had implemented a control process which would reduce and maintain the PC/M backlog to an acceptable level.

This inspection supports the position that the plant design and/or as-built configuration and performance parameters have been maintained consistent with the design bases and that design change control process activities are adequate.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 59 3.7 NRC Integrated Inspections During 1996, the NRC performed several integrated inspections at St. Lucie. Inspection Reports 96-01, 96-04, 96-06, 96-08 and 96-09 document inspection reviews of the FSAR, plant configuration, plant procedures, plant practices, and plant parameters for consistency. The general conclusions from these reports are similar to the FPL findings concerning the FSAR. The design bases information is correct but there exists minor deficiencies between the FSAR and the plant documentation. The noted deficiencies have been entered into the plant corrective action program for resolution by the issuance of Condition Reports (CR).

This inspection effort supports the position that the St. Lucie design bases and licensed configuration and performance is maintained consistent in plant areas inspected.

3.8 NRC Equipment Environmental Qualification Program Inspection In February and March 1989, the NRC conducted an inspection of the 10 CFR 50.49 requirements for environmental qualification (EQ) of electrical equipment at St. Lucie Plant. The inspection included a review of the FPL implementation program to meet the requirements of 10 CFR 50.49, walkdown inspections of EQ equipment inside containment, review of EQ maintenance activities, review of EQ design changes and FPL actions in response to NRC initiatives (NRC Notices 86-71 and 88-89), and follow-up on FPL actions in response to license conditions.

The results of the inspection were documented in NRC Inspection Report 89-07 dated April 6, 1989. No violations or deviations were identified. The results of this inspection supported the NRC's previous assessment in March 1986, that FPL had implemented an adequate program and that the program continues to be adequate. The walkdowns of EQ equipment resulted in no open or unresolved items. EQ documentation files were considered well organized and complete, requiring minimal additional information to support the analysis provided in the file. EQ m'aintenance program procedural compliance was good, and the special maintenance requirements described in the EQ document packages were confirmed to have been incorporated into the procedures for those sample items reviewed.

The inspection team noted that the tracking mechanism used by the plant to ensure scheduling and completion of EQ maintenance tasks appeared to be fragmented with separate lists for calibration, surveillance, preventive maintenance, and replacement, There was no overall program to ensure that maintenance activities were being accomplished. This was considered a weakness in FPL's EQ maintenance program. FPL acknowledged the concern and committed to implement a formalized maintenance tracking system as outlined in the inspection report details.

This inspection supports the FPL's position that St. Lucie equipment under the EQ Program was configured, qualified, performance tested, and maintained in accordance with the plant design bases.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 60 3.9 NRC Inspection of the Motor Operated Valve Program The FPL MOV program involved a design basis review and plant configuration verification by engineering and plant maintenance to determine parameters critical for MOV operation in response to GL 89-10, Third party reviews by contractors and INPO provided guidance on methodology for testing and lessons learned from other utility programs. Some of the aspects of this program include the incorporation of MOV data, setpoints, and parameters into engineering drawings, performance testing, and plant modifications to improve the operation/reliability of MOVs. The incorporation of the MOV data into the engineering process adds process controls which maintains consistency between the MOV design basis and the plant configuration.

In September 1991, the NRC conducted a special announced team inspection to examine the program developed in response to NRC Generic Letter (GL) 89-10, "Safety-Related Motor Operated Valve Testing And Surveillance." The results of the inspection were documented in NRC Inspection Report 91-18 dated November 18, 1991, and found that the basic program adequately addressed most of the recommendations. However, NRC concerns resulting from the inspection included some potential deviations from the recommendations of the generic letter, a vendor recommendation that was not being met, and programmatic details that were not sufficiently defined to assess at the time of the inspection. Two potential program deviations were resolved with the NRC in subsequent correspondence.

The NRC inspection team also noted that the personnel involved in the program were knowledgeable of the issues involved in the generic letter. Good refresher training was being provided for the upcoming October 1991 refueling outage. FPL recognized that standard industry valve factors are not wholly adequate for gate valve setting calculations. Positive steps had been undertaken to address industry concerns at the time regarding the capabilities of the diagnostic testing equipment previously used. Preventive maintenance procedures provided definitive information regarding the attributes to be assessed.

3.10 NRC Check Valve Inspection In January 1993, the NRC conducted a routine announced inspection in the areas of licensee programs associated with safety related check valves. The results of the inspection were documented in NRC Inspection Report 93-01, dated February 23, 1993. The report concluded that FPL had implemented a satisfactory check valve program to ensure the operability of check valves. They noted that knowledgeable and experienced personnel were involved in the program to ensure adequate actions were taken to address check valve problems when identified, and that there was adequate management attention to the check valve program. The NRC found that FPL's involvement with the Nuclear Industry Check Valve Users Group was a positive initiative, that check valves in the IST program were included in the check valve program, that check valves in systems beyond those identified in INPO SER 86-03 were included in the check valve program, and the use of a check valve database to identify individual valve parameters and summarize valve

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 61 maintenance history was a strength. One non-cited violation and one unresolved item were identified in the report. (Corrective actions associated with this inspection have been completed.)

This inspection supports the FPL position that St. Lucie implemented a satisfactory check valve program to ensure operability of check valves.

3.11 NRC Architect/Engineer Team Inspection From November 4, 1996 to January 10, 1997, the NRC conducted a special announced team inspection in the area of safety system operational performance capability. This particular inspection is part of a new NRC inspection initiative, called an Architect/Engineer (A/E) Team Inspection. The NRC inspection team consisted of an NRC team leader and 5 A/E inspectors (2 mechanical engineers, 1 instrumentation and controls engineer, 1 electrical engineer and 1 field engineer). The NRC team utilized the guidance of NRC Inspection Procedure 93801, "Safety System Functional Inspection (SSFI)," during the inspection with the emphasis on engineering d'esign.

The primary objective of this inspection was to assess the operational performance capability of selected safety systems through a detailed engineering design review. Two safety systems were selected for review by the NRC: the Unit 1 auxiliary feedwater (AFW) system, and the Unit 2 component cooling water (CCW) system. The inspection focused on identifying the design basis functions of the selected systems through a detailed review of license and design bases documentation. Confirmation that the selected systems could perform these design basis functions was performed through a review of engineering analyses/calculations, operating procedures, maintenance procedures and system/component testing.

In preparation for the inspection, FPL conducted an extensive self-assessment of the selected system design bases and design/configuration control processes. Documents reviewed during this self-assessment included calculations, original equipment specifications and purchase order correspondence, FSAR, NRC SERs, DBDs, Technical Specifications, license amendments, as-built documentation, accident analyses, vendor technical manuals, PCMs, TSAs, plant operating procedures, testing procedures, and corrective action program findings. Discrepancies and inconsistencies identified through this self-assessment were documented using the CR process.

None of the discrepancies identified through the self-assessment were determined to involve an issue that compromised operability of the selected systems or its capability to perform its intended design basis functions.

During the conduct of the inspection, the NRC inspection team performed extensive and detailed reviews in the following general areas:

~ Engineering design and configuration control

~ Consistency in design basis documents (FSAR, DBD, etc.)

~ System modification history and associated 10 CFR 50.59 safety evaluations

~ Design interfaces with operations and maintenance

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 62

~ Adequacy of operating procedures

~ Adequacy of Technical Specification surveillance testing for operability determinations

~ Corrective action history The preliminary NRC findings indicated that there were no major program weaknesses or failures, and that the systems inspected were capable of performing their intended design basis functions.

The NRC inspection final report is scheduled to be issued in late February 1997.

3.12 FPL QA Audits and Reviews Numerous formal QA audits and reviews have been performed by the Quality Assurance organization since the initial startup of the units, in areas related to the control of design, configuration, and testirig of the St. Lucie Plant. Design program audits typically reviewed engineering processes, implementing procedures, and the implementation of design changes and associated records (i.e. modification packages). QA audits consistently found that Quality Assurance Program elements (including design basis requirements) were adequately addressed by procedures, and that the implementation of those procedures was effective. Additionally, vertical slice reviews and safety system walkdowns were performed on a sampling basis to verify that the existing plant configuration matched the design. QA consistently found that plant safety systems were built as designed, with minor exceptions. Minor discrepancies were identified and formally resolved through the corrective action program.

Examples of QA audits and self-assessments which evaluated whether the St. Lucie systems, structures, component configuration and performance were consistent with the design bases are listed in Appendix B, Table C 3.12. Audit Findings have resulted in enhancements to engineering, operations, and testing processes and procedures.

~ QSL-OPS-88-595, "Vertical Slice Audit of the Unit 2 Intake Cooling Water System."

(5 findings - No findings involved discrepancies between as-built conditions, Inservice Testing (IST) system performance and the system design bases).

~ QSL-OPS-88-672, "Verification of the Unit 2 Reactor Coolant System Design, Operation and Maintenance, to Assure System Operability." (1 finding - regarding instrument calibration - no configuration, performance, design basis discrepancies).

~ QSL-OPS-90-739, "Verification of the Unit 1 and 2 Auxiliary Feedwater System Design, Operation and Maintenance, to Assure System Operability." (1 finding-regarding an instrument root isolation valve position - no configuration, performance, design basis discrepancies).

~ QSL-OPS-91-821, "Verification of the Unit 1 Reactor Protection System Modifications, Testing, Maintenance, and Procurement, to Assure System Operability." (No findings).

St. Lucio Utiits l and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 63

~ ISEG-PSL-A93-017, "ISEG Assessment Activities Associated With the Spring 1993 St. Lucie Unit 1 Refueling Outage." Performed to ensure that the key safety functions for shutdown conditions (i.e. RCS inventory, decay heat removal, power availability, reactivity control, and containment integrity) were not compromised by outage work.

(No findings regarding configuration, performance, design basis).

~ ISEG-PSL-A93-019, "ISEG Assessment Activities Associated With the Spring 1993 St. Lucie Unit 1 Refueling Outage." Performed reviews of the implementation of shutdown margin verifications, and pressurizer safety relief valve modifications. (No findings).

~ ISEG-PSL-A93-020, "ISEG Activities Associated With the Spring 1993 St. Lucie-1 Refueling Outage." Independent oversight of the AC and DC power availability during the outage, ensuring applicable requirements were satisfied. (No findings regarding configuration, performance, design basis). I

~ ITR 95-002, "Review of LER 335/94-006; Containment Integrity Outside of FSAR Assumptions Under Limited Circumstances due to Design Error." Evaluated the root cause of the event which involved a design error made during the 1978 addition of the iodine removal system for Unit 1. (No findings).

~ ITR 96-013, "Safety System Walkdown (Partial): Auxiliary Feedwater System, Unit 1." Verified that plant drawings reflect as-built conditions. (No findings).

II

~ QR 96-0006, "Fourth Quarter 1996 ESF Walkdown of the Unit 2 High Pressure Injection System." Verified as-built conditions are reflected in plant drawings and met UFSAR requirements, Valve lineups were performed using operating procedures, and components were verified to match design documentation. (No findings).

Other QA audits and assessments listed in Appendix B, Table 3.12 resulted in findings which directly related to St. Lucie system, structure and component configuration and performance consistency with the design bases. Resolution of these findings from the following audits listed in Appendix B, Table C 3.12 resulted in improvement in the St. Lucie configuration and performance consistent with the design bases:

~ QSL-OPS-88-624, "Unit 1 and 2 Emergency Diesel Generators and Safety-Related Switchgear Systems." (5 findings - regarding drawing/as-built discrepancies, IST discrepancies, and design basis component sizing documentation; discrepancies were resolved through the corrective action program).

~ QSL-OPS-91-800, Supplement 1, "Vertical Slice Audit of the Unit 1 and 2 Feedwater and Main Steam Systems." (2 findings - regarding drawing/as-built discrepancies and procedures/as-built discrepancies; discrepancies were resolved through the corrective action program).

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 64

~ ITR 96-015, "Safety System Walkdown (Partial): Sections of Emergency Core Cooling Subsystems in the Unit 1 Reactor Containment Building, and Pressurizer Code Safety Valve Modifications." (3 findings - regarding update of design basis documents following design modification implementation; discrepancies were resolved through the corrective action program).

~ QSL-PM-96-18, "September 1996 PSL QA Performance Monitoring Audit."

Performance-based observations in the areas of operations, maintenance, services, and engineering. (6 findings - regarding drawing/as-built discrepancies, procedures/as-built discrepancies, update of design basis documents following design modification implementation, and 2 findings that were indicative of less than adequate implementation of 10 CFR 50.59 requirements; discrepancies were resolved through the corrective action program).

4.0 Conclusion There is adequate rationale for concluding with reasonable assurance that the systems, structures and components are consistent with the design, and that discrepancies between design documents, the FSAR, procedures, and the plant are not significant with respect to safety. This conclusion is based upon following:

~ The verification at initial licensing that the plant, FSAR and the design are consistent.

~ Processes are in place to ensure the maintenance of that consistency.

~ Recent programs and initiatives have aided in verifying and enhancing the design bases for configuration and performance requirements on plant systems.

~ Audits and inspections have verified consistency between the design basis and the plant.

~ Processes have been and are in place for identification, evaluation and correction of discrepancies between the design basis and the plant systems, structures and components.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 65

[d] "Processes for identification of problems and implementation of corrective actions, including actions to determine the extent of problems, action to prevent recurrence and reporting to NRC."

The following outline provides the organization of the response to request [d]:

1.0 Introduction - Background 2.0 Problem Identification Processes 2.1 Operator Rounds 2.2 Walkdowns 2.3 Operating Experience Feedback (OEF) /Generic Implications 2.4 Management Self-assessment 2.5 Quality Assurance (QA) Audit Programs 2.6 Nuclear Safety Speakout (SPEAKOUT) 3.0 Corrective Action Programs 3.1 Historical Discussion 3.2 Quality Instructions (Qi) 3.3 Condition Reporting (CR) System 3.4 Event Response Teams 3,5 Commitment and Corrective Actions Tracking (CIRAC) Systems 3.6 Change Request Notices (CRN) 3.7 Nonconformance Reports (NCR) 3.8 10 CFR Part 21/Significant Safety Hazard Evaluations 3.9 Operability Determination 3.10 10 CFR 50.72 and 50.73 Reportability Determination 3.11 Corrective Action Program Training 4.0 NRC / St. Lucie Interfaces

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 66 1.0,. Introduction - Background The Site Vice President and Director of Nuclear Assurance are responsible for the overall implementation of the programs described in this response. Operations personnel are responsible for operating the plant in accordance with the design bases, Technical Specifications and the applicable regulations.

Along with a questioning attitude exhibited by plant personnel and the identification of issues during routine work, St. Lucie has many programs which are specifically designed to identify issues needing resolution. Examples which are discussed in this response are, operator rounds, programinatic and special walkdowns, Operating Experience Feedback, management self-assessments, the QA program, the Condition Report process, and Nuclear Safety SPEAKOUT.

Conditions found may be documented by any employee, permanent or temporary, on a Condition Report (CR). After review by the Nuclear Plant Supervisor and the Plant General Manager (PGM), the organization assigned responsibility for resolution of the condition conducts an analysis and develops appropriate corrective actions. The corrective actions identified in the resolution to each CR are tracked by personnel responsible for completion of the assigned activities. The CR program is administered by the Services organization.

Root causes are typically determined by personnel from the organization assigned responsibility for resolution of the CR. The depth of analysis and resources devoted to corrective action is commensurate with the significance of the noted condition.

The Nuclear Plant Supervisor and the on-shift operations staff are authorized and required to make timely reportability and operability decisions and follow through the applicable procedures to completion. For those conditions which are of a nature which allow or require more detailed evaluation, the Engineering staff aids in the determination of operability and the Licensing staff aids in the determination of reportability.

2.0 Problem Identification Processes 2.1 Operator Rounds Operator rounds are required by plant procedures. The procedure in some cases implements the requirements of Technical Specifications and is also a reflection of management expectations for the operator to keep abreast of the operation of the equipment in the plant.

Operator rounds use a computerized data logger to provide the operator with detailed plant information. This information is useful in documenting plant status and is more user-friendly than the previous hard copy logs. The computer provides the operator with the range of normal values associated with equipment performance. It also has Technical Specification information, trending capabilities, and does not allow the operator to exit the program until the rounds have been

St. Lucie Uqits 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 67 completed. The availability of the information in the computerized data set simplifies the identification of an abnormal condition and provides a more efficient means for supervisor review.

Plant procedures provide broad scope requirements, as well as specific departmental instructions governing the aspects of each watch station. Satisfactory completion of a set of rounds on a particular watch station in accordance with these instructions (under the surveillance of a supervisor) is a requirement for qualifying on a watch station. The instructions are periodically used as a monitoring tool by management to insure that the rounds are performed in accordance with their expectations. The rounds help to trend and respond to degrading conditions in a timely manner. Plant work orders are generated for deficiencies, and condition reports are generated for items outside the work order process.

2.2 Walkdowns System engineers and component engineers are required to perform walkdowns of their systems on a regular basis. Directions for walkdowns are provided by procedure. Engineering Department instructions are in place for walkdowns of components and systems. See Section 4.0 in the response to question [c] for more details on programmatic and special walkdowns. If nonconforming conditions are found, the CR system is used to document the condition and provide a vehicle for the correction of the condition.

~

2.3

~ Operating Experience Feedback (OEF) /Generic Implications Industry events and reports are reviewed and processed based on applicability to St. Lucie in accordance with administrative procedures. Industry events and reports are reviewed by the OEF coordinator and processed based on potential generic applicability to St. Lucie in accordance with Administrative Procedure ADM-17.03, "Operating Experience Feedback." Industry event reports screened by the OEF Coordinator include:

NRC Generic Letters (GL)

NRC Information Notices (IN)

NRC Daily Event Reports NRC Press Releases (PR)

INPO Significant Operating Experience Reports (SOER)

INFO Significant Event Reports (SER)

INPO Significant Event Notifications (SEN)

INPO Significant by Others Reports (SO)

INPO Operations and Maintenance Reminders (0&MR)

INPO Operating Experience Reports (OE)

Westinghouse, General Electric, and Combustion Engineering Vendor and Technical Reports

~ ~ ~ ~

Operating experience items which merit analysis, specific actions or a thorough, documented plant

~

review or response, are processed as CRs. Implementation of the CR process ensures that a

~

~

thorough evaluation and analysis of the event or report is made and that the plant's response (if

St. Lucie Ututs 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 68 applicable) adequately addresses the concerns. Ifan update to the plant's design bases documents is tcquired in response to the operating event, it willbe identified and implemented through the CR process using the appropriate engineering processes.

Where enhancements or changes are identified which would benefit the plant or mitigate the consequences of an event, procedure/process changes are made as necessary. One of the OEF program goals is to aid in the development of plant procedures, such that the potential for the introduction of a step which could initiate an event is minimized during procedure change process.

Violations which have been cited at other licensees are reviewed on a periodic basis for applicability to St. Lucie. Contractors that have compiled docketed information on internal as well as generic communications from the NRC have been retained by FPL. This body of information is available in an electronically searchable format, and aids in the timely review of information potentially applicable to St. Lucie. Recent access via the Internet has made many of these documents available to FPL on the same day as released to the industry by the NRC.

2.4 Management Self-assessment St. Lucie departments have a series of formal and informal programs which provide for self-assessment.

The self-assessment processes used at St. Lucie include:

Post Outage Assessments Post Trip Reviews Off Hours Tours Training Observations Plant Manager's Walkdown Event Response Teams Condition Report System Trending Quarterly Trend Analysis Each of these processes provide the opportunity to identify procedure, design or configuration issues in need of resolution.

Beginning in August 1995, a series of significant problems and events indicated that St. Lucie plant performance had declined. St. Lucie plant management took the initiative to conduct a critical and comprehensive self-assessment of St. Lucie Operations, Engineering, Maintenance and Plant Support functional areas, as well as management policies and the corrective action program.

With respect to the St. Lucie design bases, the assessment did not identify any safety significant issues, However, some opportunities for improvement were identified in related areas such as configuration management. For example, the accountabilities for the maintenance of the FSAR was not clearly established, some plant drawings, procedures and licensing documents were not updated to reQect modifications, and there was insufficient training provided to engineers assigned

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 69 to identify and perform required drawing updates. Corrective actions have been initiated for the weaknesses identified by the self-assessment.

2.5 Quality Assurance (QA) Audit Programs The QA organization at St. Lucie is an independent organization reporting through an offsite Director directly to the President of the Nuclear Division. QA has a program of audits which is required by Technical Specifications. Audits are also conducted at the request of the CNRB and upper Nuclear Division management. The findings and recommended corrective actions resulting from the audit program are documented through the generation of a CR for each item, as well as in the QA audit files. (Also see discussion in the response to request [a], Section 6.0, on the Topical Quality Assurance Report (TQAR) Section 18.0 requirements for audits.)

2.6 Nuclear Safety Speakout (SPEAKOUT)

The SPEAKOUT program is a separate confidential reporting process for identification of concerns. This program may be used by personnel wishing to remain anonymous or used as a method outside the normal chain of command reporting process. This program is designed to maintain anonymity while seeking causes and corrective actions for the identified issues. A designated committee reviews significant SPEAKOUT issues for thoroughness of investigation, documentation and corrective actions. Summary reports on each issue are returned to the identifying individual for their information.

3.0 Corrective Action Programs 3.1 Historical Discussion The overall method and process for addressing corrective action, as required by Criterion XVIof 10 CFR 50 Appendix B, has historically been a controlled process, with engineering support provided as necessary. Issues are raised through successive management levels, up to and including the President Nuclear Division. This process is progressive, and the individual may continue up this chain ifsatisfactory action is not taken at any level on an individual concern. This process is supplemented by the Nuclear Safety SPEAKOUT program which provides a mechanism for anonymously identifying concerns. See section 2.6 of this response for further details on SPEAKOUT.

The TQAR for St. Lucie is the top tier document (approved by the NRC) which defines the QA program. The TQAR is written such that the numbering and topic of each section matches that of 10 CFR 50, Appendix B. Quality Instructions and plant administrative procedures provide implementation level guidance for TQAR requirements. This hierarchy of quality procedures has existed since the nuclear unit construction time frame. The basic requirements for corrective action processes are described in the following Topical Quality Reports (TQR):

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 70 This TQR requires that measures be established to control nonconforming items to prevent them from being installed or used. Physical controls such as segregation or tagging ensure that inadvertent use of nonconforming items does not occur. Measures are also provided for the evaluation and disposition of the items. For example, nonconformances which affect design integrity or create a deviation from design specifications require Engineering review for identification of appropriate corrective actions. A documented technical evaluation is required when an item is accepted "as-is" or repaired to an acceptable evaluation. These evaluations assure that the final condition does not adversely affect safety, operability or maintainability of the item or component.

This TQR requires that measures be established to ensure that conditions adverse to quality are promptly identified, tracked and corrected. These measures currently exist in the form of the Condition Reporting system described below. Ifthe identified condition is significant, a Root Cause Analysis is performed, the extent of the condition is determined (generic implications),

appropriate actions are taken to prevent recurrence, and results are reported to appropriate management levels. Conditions which may affect plant system operability receive a documented operability assessment within specified time frames. Requirements have been in place to ensure that corrective action not only corrects the immediate condition but also prevents adverse conditions from recurring.

This TQR requires the conducting of a comprehensive system of planned and periodic audits of the quality assurance program to determine the effectiveness of the program. Audit results are reviewed by the management of the audited organization and subsequently by the Company Nuclear Review Board.

3.2 Quality Instructions (QI)

Quality Instructions are procedures written to delineate actions and organizational responsibilities which affect quality in the operation of the St. Lucie Plant. Quality Instructions provide implementation level detail for the Quality Assurance Program requirements outlined in the TQR.

Similar to the TQR, the set of Quality Instructions also follows the criterion structure of 10 CFR 50 Appendix B.

Quality Instructions within the QI 15 and QI 16 series have historically defined the methods for requesting, implementing, documenting and reporting actions to correct identified deficiencies.

These procedures provide for disposition by Engineering of nonconformances affecting design integrity or deviations from design specifications. Provisions have been included to address concerns of an immediate nature, to track open items until closure and to provide for reporting of significant conditions to management.

St. Lucie Ututs 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 71 3.3 Condition Reporting (CR) System Prior to 1994, St. Lucie had several problem reporting and resolution systems. In 1994, most were consolidated into a single site process.

The CR process is controlled by an administrative procedure (AP0006130, "Condition Reports" ).

The process is used throughout the Nuclear Division, although each plant has its own implementing procedure. The system is also designed to have a very low threshold for use by anyone working in the Nuclear Division. This allows tracking and trending of low level as well as high level issues to aid plant management to set priorities on issues as they arise.

The CR process includes requirements that each CR be reviewed by the Nuclear Plant Supervisor (NPS) for the purpose of maintaining his awareness of plant conditions and the performance of an initial operability assessment. Subsequent to this review by the NPS, condition reports receive a review by an interdisciplinary panel of plant personnel. In addition, this panel makes recommendations to the Plant General Manager regarding the level of root cause analysis required and the appropriate department for assignment. Subsequent to this review, the Plant General Manager reviews all Condition Reports. The CR process includes the following corrective action evaluations:

1. CRs are screened for safety significance, operability and reportability issues (10 CFR 50.72 - Notification Requirements; 10 CFR 50.73 - LERs, and 10 CFR 21) and are prioritized according to their potential impact on continued safe operation.

2, For CRs of safety significance, operability, and reportability evaluations are initiated immediately following initial screening in the corrective action process.

3. Root cause analysis and review for generic implications is performed commensurate with the significance of the condition.

4. Corrective actions are developed, reviewed by management of the assigned department and implemented.

5. Corrective action implementation is scheduled according to the prioritization of the issue (including consideration of impact on design bases documentation).

6. Condition Reports receive an independent review for closure.

7. Tracking and trending of CRs is performed on a periodic basis for the purpose of identifying performance trends and problem areas requiring management attention.

Corrective actions which are generated by the CR process are tracked in a Plant Manager Action Item (PMAI) tracking system. This is a plant specific tracking system which follows the corrective actions to completion and gives the items visibility within the organization.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 72 3.4 Event Response Teams An Event Response Team (ERT) program provides for a fast acting group of knowledgeable plant personnel to investigate the root cause of plant events such as, automatic reactor trips, unplanned unit shutdowns, and recurring maintenance/repair issues, Significant equipment problems can also be the subject of investigations by an ERT. The ERT process is defined within an ERT Guideline document. An ERT may be requested by any member of the Nuclear Management Team. The PGM or Engineering Manager appoints an ERT leader and concurs with the selection of team members. In cases where an ERT is formed, a CR is initiated.

Team meetings are held as often as necessary until the ERT accomplishes the following tasks:

~ Determination of the root cause of the event.

~ Determination of short-term countermeasures.

~ Determination of countermeasures to prevent recurrence of the event.

~ Identification of possible generic problems that may cause similar events, and development of action plans to address them.

The ERT uses tools such as event time lines, drawings, cause and effect diagrams, failure modes and effect analysis, possible root cause evaluation matrices, system operators, vendor application experts, system engineers, and component specialists to aid in the determination of root cause. A formal plant procedure, ADM-08.04, "Root Cause Evaluation," is also in place to provide detailed guidance in determination of zoot cause. This procedure provides for use of the following proven root cause techniques: task analysis, change analysis, barrier analysis, event time line and causal factor charting, causal factors category listing, cause and effect analysis, and fault tree analysis; A formal procedure, OP0030119, "Post Trip Review," is in place to provide a systematic method for ensuring the proper functioning of safety-related and other important equipment during a reactor trip. The post trip review is used as input to the ERT and provides management with the necessary information to determine when the plant can be safely restarted.

Event Response Team recommendations and countermeasures are documented and tracked using the CR system.

3.5 Commitment and Corrective Actions Tracking (CTRAC) Systems In 1985 St. Lucie began tracking commitments made to the NRC. The tracking goal is to meet the completion date for all commitments. Commitment to Excellence Program (CEP) Project 6 was implemented to enhance the effectiveness of licensing activities and to improve the coordination and control of NRC-related projects. This tracking system provided a process for closing each commitment item, such as those made within an application for a license amendment, an exemption request, a response to a notice of violation, a corrective action within a Licensee

St. Lucie Uttits 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 73 Event Report, or other commitment made to the NRC on the docket. The commitment tracking system today is more detailed but essentially the same in function as that generated as part of Project 6. NRC commitments are now tracked using the PMAI Tracking System.

An additional feature of the system is to make available a search tool which can ascertain ifa particular issue is in fact a commitment and provide a link to the originating document.

Procedures often have commitment references in the introductory sections of the procedure and in many cases in the body of the procedure next to the step which implements the commitment.

3.6 Change Request Notices (CRN)

When an engineering design package is issued, there is always a potential that some field condition or minor change may be required after the package is issued. The Engineering procedures permit a design package to either go through a formal revision or to have a CRN be issued to resolve the minor change. The CRN is the primary method used to make minor changes to an issued engineering package, where the requested change does not affect the design or safety analysis of the engineering package. A CRN may also be initiated for a design package by any individual and provides a mechanism for formally correcting minor potential problems with an issued design.

3.7 Nonconformance Reports (NCR)

The NCR process was administered by the Quality Control Department until 1994, at which time, it was made part of the Condition Report (CR) process which is now administered by the Services Department.

The NCR process provides a means for controlling the disposition of nonconforming items reported during receipt inspection or field inspection of SSCs where Engineering is required to evaluate and resolve the nonconformance. The NCR process also provides for the conduct of an operability assessment for nonconformances identified on in-service equipment. Nonconforming items that are acceptable by virtue of existing design documentation do not require Engineering evaluation and are corrected by a Plant Work Order or Relay Work Order.

3.8 10 CFR Part 21/Significant Safety Hazard Evaluations One mechanism for evaluating design or equipment problems is through the 10 CFR 21 reporting process, When first instituted in 1977, Part 21 was used extensively to address design and equipment issues identified both internally and by external vendors. 10 CFR 21 evaluations have almost exclusively been performed by engineering personnel ifthe design or equipment is found to have been provided to FPL. These evaluations provided a means for FPL to evaluate and address specific issues from either external or internal sources and to document required corrective actions and reporting to NRC. Subsequent changes in this rule and the promulgation of 10 CFR 50.72 and 50.73 are such that evaluations are now primarily performed with "operability" criteria as the focus rather than against substantial safety hazards criteria. Operability is incorporated

St. Lucie Uttits 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 74 within the condition reporting process and is more restrictive than 10 CFR 21. 10 CFR 21 evaluations are performed for equipment received by the warehouse, but not yet installed in the power plant to evaluate ifa substantial safety hazard exists. The reporting criteria of 10 CFR 50.72 and 50.73 and their use will be discussed in section 3.10.

3.9 Operability Determination The initial operability review for all conditions is conducted during the Nuclear Plant Supervisor's review of the Condition Report. Ifa more detailed review is warranted, the CR is assigned to Engineering. Recommended guidance, considerations, and methodology for performing assessments of operability on nonconforming or degraded conditions are proceduralized in a Nuclear Engineering QI. The QI is based on the information contained in NRC Generic Letter 91-18, "Information to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability."

The following aspects of operability determination are discussed in the QI:

~ Operability vs. full qualification

~ Operability vs, single failure

~ Operability and the use of manual action vs. automatic action

~ Operability and the use of probabilistic safety assessment

~ Operability vs. environmental qualification

~ Technical Specification operability vs. ASME Section XI operative criteria

~ Operability vs. piping and pipe support requirements

~ Operability vs. flaw evaluation

~ Operability vs. operational leakage (Generic Letter 90-05)

~ Operability vs. structural requirements In addition, a design/operability reference guide is in the Engineering QIs to provide additional guidance on operability considerations as well as design, safety, quality, and regulatory considerations which may facilitate development of thorough design inputs/analyses and operability assessments. The guidance is intended to be a checklist to ensure that applicable considerations are not omitted during a design or operability assessment activity.

3.10 10 CFR 50.72 and 50.73 Reportability Determination Conditions which may be reportable are documented on a Condition Report. Each CR is reviewed by the Nuclear Plant Supervisor for 10 CFR 50.72 reportability in accordance with administrative procedure AP 0010721, "NRC Required Non-Routine Notifications and Reports."

The site Facility Review Group (FRG) reviews specific events for reportability and the Licensing Department prepares all Licensee Event Reports (LER).

St. Lucio Ututs 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 75 3.11 Corrective Action Program Training Training on identification of problems and implementation of corrective actions includes the following:

The Engineering Support Personnel (ESP) Training Plan includes initial training on the Corrective Action Program. This training is part of the Initial Indoctrination training section of the INPO accredited Engineering Support Personnel (ESP) Training Program. This training consists of a basic review of the plant procedures associated with the corrective actions program including the CR process. Specific training in root cause analytical techniques has been periodically provided to St. Lucie personnel. This material is contained in a handbook entitled "Problem Identification and Correction" (PIC). This handbook includes topics such as, defining the problem, determining root cause and the development of appropriate corrective actions. The PIC process training is offered to appropriate management and engineering personnel.

Methods available for the reporting of problems at St. Lucie is a subject of the General Employee Training Program presented to plant employees and contractors on an annual basis. Additionally, the opportunity to provide information directly to the NRC in accordance with the provisions of Form 3 is a subject of this trauung.

"Notification of Plant Events" is a lesson plan which includes the reporting requirements of 10 CFR 50.72. and is used for the training of licensed operators 4.0 NRC/St. Lucie Interfaces FPL has a policy that requires open and candid communications with NRC personnel. Frequent communications are made with the NRC resident inspectors and the NRC project manager for St.

Lucie. A regulatory liaison office is maintained in Washington D.C. to allow expedited face-to-face communications with the NRC staff, ifrequired, for any issue which may arise,

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 76

[e] "The overall effectiveness of your current processes and programs in concluding that the configuration of your plant(s) is consistent with the design bases."

An outline of the organization of the response to this request follows:

1.0 Introduction 2.0 Program Reviews Which Lead to Design Bases Verification 2.1 NRC Bulletin 79-14 Piping Walkdown Program 2.2 Technical Specification Improvements 2.3 Electrical Power System Design Bases Enhancements 2.4 Updated LOCA Containment Analysis 2.5 St. Lucie Unit 1 Thermal Power Uprate 2.6 Containment Penetration Boundary Review 2.7 St. Lucie Unit 1 - USI A-46 Seismic Adequacy Program 2.8 St. Lucie Unit 1 Steam Generator Replacement Project (SGRP) 2.9 FSAR Procedural Consistency Review 3.0 FPL Vertical Slice Audits and Conclusions 3.1 Vertical Slice Audit of the St. Lucie Unit 2 Intake Cooling Water System (ICW) 3.2 Vertical Slice Audit of the Units 1 and 2 Feedwater and Main Steam Systems 3.3 Quality Assurance Audit Conclusions 4.0 NRC Vertical Slice Audits 4.1 Electrical Distribution System Functional Inspection (EDSFI) 4.2 Service Water System Operational Performance Inspection (SWSOPI) 4.3 1996-1997 Architect/Engineer Inspection 5.0 Conclusions on the Overall Effectiveness of the FPL Program at St. Lucie

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 77 1.0 Introduction FPL, with reasonable assurance, is confident that the current processes and programs are effective in ensuring that the St. Lucie design and configuration are consistent with the design bases. This conclusion is based upon the following:

The initial plant verification process, startup testing, surveillance testing, and control of the modification process results in consistency between plant design and procedures. In addition, previously performed design bases verification efforts such as SSFI-type inspections, the FSAR procedural consistency review, and related reviews show the consistency between the plant, plant procedures, and the design bases.

2. The plant design change process ensures the compatibility of the plant physical and functional characteristics with the design bases and plant documentation. Changes are evaluated and reviewed to fully determine the impact of each change on other systems and documents and ensure adherence to established requirements.

3. The processes for identifying, analyzing, and resolving conditions include the Condition Report process (which includes the review of potential defects and noncompliances and Engineering review of selected CRs), the Nuclear Safety SPEAKOUT program and the QA audit program. For cases where problems were and are identified, appropriate root cause analyses and corrective actions were and are implemented.

Based upon information presented in the responses to NRC requests and the items presented above, FPL concludes that its design control processes provide reasonable assurance and confidence that the plant is being operated and maintained within its design bases in a manner that does not adversely impact the health and safety of the public.

The design control process integrates plant changes and modifications into the plant's operating, maintenance, and testing procedures and assures that the plant's performance is consistent with its design bases.

FPL has assessed the accessibility and adequacy of its design bases documentation. The results of this assessment form the basis for the FPL conclusion that sufficient design bases reconstitution has been accomplished to provide an adequate baseline for timely operability determinations and plant design changes. The design bases are sufficiently understood, documented, and auditable.

Operability determinations and 10 CFR 50.59 evaluations that may need to be made quickly in response to plant issues can be supported. The St. Lucie record for complete and accurate operability determinations supports this overall conclusion regarding design bases adequacy.

Information used solely to support development of design modification packages need not be retrieved as expeditiously but is readily retrievable as evidenced by the preparation for and results of the St. Lucie NRC A/E Inspection.

St. Lucie Ututs 1 and 2 Docket Nos. 50-335 and 50-389 Qe L-97-28 Enclosure Page 78 The FPL methodology used in design bases reconstitution decisions considers the safety significance of any missing or erroneous information. The FSAR procedural consistency review, the NRC A/E Inspection preparation effort, and the NRC A/E Inspection results identified discrepancies among design bases information resources, none of which constituted a significant safety concern. The NRC's A/E Inspection performed at St. Lucie found that the design bases changes for the systems inspected had been adequately implemented. These assessments and inspections did not identify programmatic failures to adequately maintain the plant's design bases.

FPL's confidence is based on the evaluation of a number of factors and prog'rams discussed in previous sections of this enclosure. Design and configuration control programs described in NRC Request [a] and the results of evaluations described in response to NRC Requests [b] and [c]

support our conclusion that the design and configuration control processes used at St, Lucie are in accordance with regulatory expectations. These programs have been improved over time as industry and company expectations have changed. These program improvements have culminated in effective design control and configuration management and have correctly translated design bases information into the plant operating, maintenance and testing procedures.

Specific activities which validate the overall effectiveness of maintaining St. Lucie's design bases information are discussed below.

2.0 Program Reviews Which Lead to Design Bases Verification 2.'1 NRC Bulletin 79-14 Piping Walkdown Program NRC Bulletin 79-14 "Seismic Analysis for As-Built Safety-Related Piping Systems" addresses the concern that the seismic analyses of safety-related piping, coupled with pressure, thermal, operating weight, and other applicable system operating parameters, be verified in accordance with actual as-built drawings for its impact on the stress analysis of record. At St. Lucie Units 1 and 2, the verification process consisted of a formal program by which as-built configurations of Seismic Category I safety-related piping systems were documented, evaluated, and modified as necessary under the FPL QA/QC program as appropriate, by a team of experienced engineers and designers.

This program consisted of walkdowns of designated Seismic Category I safety-related piping systems and documentation of the as-built configuration including piping size, components, and geometry; location, orientation, and type of valves; location and configuration of pipe supports; and floor and wall penetration details. These details were then evaluated by the team to assess if document revisions and/or plant modifications were required and, ifnecessary, acceptable under the construction code of record.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosttre Page 79 2.2 Technical Specification Improvements The St. Lucie units were licensed to the standard technical specification formats at the time the operating licenses were issued. Over the years there have been 147 amendments to the St. Lucie Unit 1 operating license and 86 amendments to the St, Lucie Unit 2 operating license, the majority of which were revisions to the units technical specifications. St. Lucie has reviewed and implemented many of the Technical Specification line item improvements issued as NRC Generic Letters (GL) and has adopted a number of the NUREG 1432, "Standard Technical Specifications, Combustion Engineering Plants," improvements.

2.3 Electrical Power System Design Bases Enhancements In 1991, FPL completed a review of the St. Lucie Electrical Distribution System (EDS). The objective of this activity was to assess the performance capability of the EDS by reviewing the design parameters as they relate to onsite and offsite emergency power sources and associated safety-related equipment which is relied on during and following design basis events.

Additionally, key parameters of plant configuration and drawings were reviewed for consistency with the design.

Upon completion of the review, several calculations were updated or revised to reflect current design standards. In addition, a number of inconsistencies were identified in the FSAR and plant configuration drawings. These inconsistencies were scheduled for correction and closed following the audit. During this review there were no significant operability concerns identified in the St. Lucie Electrical Distribution System.

2.4 Updated LOCA Containment Analysis In 1993, the LOCA containment analyses of record were updated. Prior to the update, the containment section of the FSARs included much of the original design basis sensitivity studies.

Impact of previous changes, such as an increase in the rated power level from 2560MWt to 2700 MWt, had been addressed. Typically, this information was updated with the most limiting case of record. Analyses, however, had been performed using a case-by-case approach, making portions of the FSAR unclear and outdated.

This 1993 update to the LOCA containment analysis incorporates applicable plant changes into one concise evaluation utilizing operating experience where necessary to justify the inputs.

2.5 St. Lucie Unit 1 Thermal Power Uprate In 1981, FPL provided analyses to justify an increased thermal power rating for Unit 1 from 2560 MWt to 2700 MWt. This resulted in numerous changes to the plant's design bases. These changes are discussed in the proposed license amendment request which was provided in

St. Lucie Uttits l and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 80 accordance with 10 CFR 50.90. This amendment request included an engineering evaluation to justify the proposed changes.

The thermal uprate project evaluations were reviewed by FPL in 1996 as part of preparations for the NRC A/E Inspection. FPL identified that the condensate storage tank requirements for auxiliary feedwater had not been updated to correspond to the approximately 5% increase in thermal power. The decay heat generated by the reactor core is increased due to the increase in thermal power. A Condition Report was generated to resolve this inconsistency. Should the analysis conclude that the condensate requirements are not bounded by the current Technical Specification minimum, then an amendment to the technical specifications willbe requested in accordance with 10 CFR 50.90.

Additionally, the 1996 FSAR procedural consistency review found several updating inconsistencies resulting from the thermal power uprate. Condition Reports were issued for resolution of these inconsistencies.

2.6 Containment Penetration Boundary Review Technical assessments of containment penetration boundaries for Units 1 and 2 were completed in 1991. The objectives of these evaluations were to perform a technical assessment of each mechanical containment penetration to determine those components whose integrity or position was vital to containment integrity, and to provide sketches identifying the containment boundary components. The evaluations show, for each individual penetration, the containment penetration boundary, including components that are vital to containment integrity.

The Unit 1 and Unit 2 evaluations concluded that in general the containment isolation system met the intent of 10 CFR 50 Appendix A, General Design Criteria, in that a double barrier is provided in each line penetrating the containment.

2.7 St. Lucie Unit 1 - USI A-46 Seismic Adequacy Program NRC Generic Letter (GL) 87-02, "Verification of Seismic Adequacy of Mechanical and Electrical Equipment in Operating Reactors," (USI A-46) required a review of St. Lucie Unit 1 equipment required to achieve and maintain hot shutdown of the plant for a period of eight hours. FPL selected safety and non-safety-related equipment for their review. This equipment was evaluated for seismic adequacy by FPL engineers and a seismic review team consisting of engineering experts in the area of seismic adequacy of equipment and equipment performance during earthquakes. The four basic requirements for the GL 87-02/USI A-46 review of St. Lucie Unit 1, which were verified through walkdowns are: the equipment seismic capacity being greater than demand, the construction adequacy of the equipment, anchorage adequacy, and no seismic spatial interactions.

St. Lucie Ututs 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 81 2.8 St. Lucie Unit 1- Steam Generator Replacement Project (SGRP)

The SGRP developed a 10 CFR 50.59 evaluation and a set of design packages that willbe used during the 1997 steam generator replacement outage. In the course of doing this design work, plant walkdowns were conducted and the FSAR, design analyses and other applicable design basis documents were reviewed, Walkdowns were conducted inside and outside containment in the areas affected by the SGRP. These walkdowns found minor pipe and HVAC support discrepancies between the design drawings and the as-found condition. Each discrepant condition was documented, evaluated and dispositioned. The accident analyses presented in Chapter 15 were evaluated to determine the effects of using the replacement steam generators (RSG). The objective of each evaluation was to demonstrate that the St. Lucie Unit 1 plant response with the RSGs would meet all NRC-approved FSAR acceptance criteria.

These evaluations of the existing plant design and licensing bases, steam generator physical interfaces, thermal-hydraulic aspects, structural supports, safety analyses, technical specifications, and off-normal and emergency procedures showed that the RSG design preserves the existing plant design and licensing bases. Therefore, use of the RSGs at St. Lucie 1 meets all relevant criteria and guidance, the existing licensing and design bases remain valid, the Technical Specifications and their bases remain valid, and no unreviewed safety questions exist.

2.9 FSAR Procedural Consistency Review FPL devoted about 8000 man-hours in 1996 to the review of the FSARs. This effort involved a comprehensive review of the FSARs with the primary focus being the identification of procedural processes included in the FSAR and their comparison to the current plant procedures to assure consistency.

The conclusion of this effort is that there were no operability issues of safety significance. The final results of some Condition Reports (CR) are still pending, however, the results to date have not posed any unresolved safety concerns. The small number of CR-related inconsistencies indicate that there does not appear to be major discrepancies between the FSARs and the operating procedures. Therefore, this program supports the overall conclusion that plant procedures and the design basis are consistent.

3.0 FPL Vertical Slice Audits and Conclusions 3.1 Vertical Slice Audit of the St. Lucie Unit 2 Intake Cooling Water System (ICW)

In 1988, the FPL QA department conducted a vertical slice audit on the Unit 2 ICW system design, procurement of parts, and system operation and maintenance, to assure system operability.

The ICW system and components were inspected to evaluate configuration control, and to determine whether they were capable of performing the safety functions required by their design bases. Plant modifications were reviewed to determine ifany unreviewed safety questions were

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 82 introduced. System walkdowns verified as-built configuration. Pre-operational testing and hydrostatic testing met the objectives of the design bases. The audit resulted in five findings, none of which involved discrepancies between as-built conditions and the design bases. The findings were formally addressed by the corrective action process and were closed.

3.2 Vertical Slice Audit of the Units 1 and 2 Feedwater and Main Steam Systems In 1991, the FPL QA department conducted a vertical slice audit on the design, operation, and maintenance of the Unit 1 and Unit 2 feedwater and main steam systems. The design bases, as described in the FSAR, were reviewed to assure that configuration control was maintained.

Safety evaluations for system modifications were reviewed to ensure unreviewed safety questions did not exist. System walkdowns verified the as-built configuration, component identification, orientation, accessability, and general plant conditions. Operating procedures, plant drawings, maintenance activities, and post maintenance testing were also reviewed. Discrepancies between plant drawings and procedures resulted in two findings. The findings were formally addressed by the corrective action process and were closed.

3.3 Quality Assurance Audit Conclusions The QA Audit Program, along with other technical reviews and assessments performed by the Quality Department or Independent Safety Engineering Group (ISEG), have consistently evaluated the implementation of plant modifications, compared system as-built conditions to plant design, and evaluated whether drawings and procedures were consistent with the plants'esign bases, with overall satisfactory results.

4.0 NRC Vertical Slice Audits 4.1 Electrical Distribution System Functional Inspection (EDSFI)

In February and March 1991, the NRC conducted a special announced team inspection in the areas of design of electrical systems and related engineering and maintenance activities. This inspection specifically focused on the Electrical Distribution System (EDS) as-built configuration, conformance to design bases requirements, and design output documents. The NRC concluded that the EDS at St. Lucie as evaluated was capable of performing its intended function under normal and accident conditions. It was also concluded that adequate controls are in place to maintain the EDS in an operable configuration.

4.2 Service Water System Operational Performance Inspection In September and October 1991, the NRC conducted a team inspection in the area of the service water operational performance at St. Lucie. This inspection specifically focused on the service water system (i.e., intake cooling water system) system as-built configuration and operational performance regarding the applicable design bases requirements. The inspection included a mechanical design review, detailed system walkdowns, and a review of system operation,

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 Qo L-97-28 Enclosure Page 83 maintenance, and surveillance procedures. It also included an assessment of quality assurance and corrective actions related to the intake cooling water gCW) system. The NRC inspection team concluded that the ICW as-built configuration would be capable of performing its intended safety function under design bases conditions.

4.3 1996-1997 Architect/Engineer Inspection From November 1996 through January 1997, the NRC conducted an Architect/Engineer Inspection of two systems at St. Lucie; the Unit 1 Auxiliary Feedwater and the Unit 2 Component Cooling Water Systems. This inspection was performed in accordance with NRC Inspection Manual Chapter 93801, Safety System Functional Inspection (SSFI) guidance.

FPL's preparation for this inspection identified 17 concerns associated with design bases, design or configuration control, or plant configuration/performance. Condition Reports were submitted for these items which entered the condition into the corrective action program for resolution. One Condition Report identified undervoltage trip devices on dc molded-case circuit-breakers which had not been tested for operation. The finding resulted in an NRC notification concerning the 125V dc swing-bus tie-breakers being in a condition which alone could have prevented fulfillment of a safety system. There were no other identified design bases inconsistencies which required NRC notification.

The NRC has not issued the inspection report at the time of this response, however, at the public exit meeting for the NRC A/E Inspection effort, the NRC noted that there was no breakdown or major failures in the configuration control process or operationally significant concerns with the design bases for the plant. The FPL self-assessment and the NRC inspection support the conclusion that the design bases information is available and adequate and that major program failures do not exist.

5.0 Conclusion on the Overall Effectiveness of the FPL Program at St. Lucie The events, inspections, audits, and review programs discussed above, provide the basis to conclude with reasonable assurance that the design bases information is adequate and available.

The corrective action program assures that identified problems or concerns are tracked, evaluated, and corrective actions implemented. The process also trends plant events to identify repetitive problems which could indicate a process/program problem. This allows corrective actions to be implemented on a range of issues.

The focus of recent FPL activities has been to reduce the number of changes occurring at its nuclear units such that the impact on the design bases is reduced. Backlogs of design changes and required equipment repairs/replacement are being reduced, resulting in improved control of the plant's configuration.

Based upon information presented in this 'section, as well as in the FPL response to requests [a]

through [d] of this enclosure, FPL has concluded that the existing control processes provide

St. Lucie Ututs 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 84 reasonable assurance and confidence that the plant is being operated and maintained within its design bases, The validity of these configuration control processes are confirmed by continued safe plant operation, self-initiated evaluations, and NRC assessments of plant performance. The events, inspections, audits, and programs discussed above provide additional basis to conclude that with reasonable assurance, the design bases information is adequate and available. The corrective action program assures that problems or concerns which are identified are tracked, evaluated and corrective actions implemented to resolve the problems or concerns.

St. Lucie Uttits 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 85

[f] "Supplemental request for information on design review/basis programs."

The following outline provides the organization for the response to this additional information request.

1.0 Nuclear Energy Institute (NEI) 96-05 Initiative 2.0 FSAR Reviews and Self-Assessments 3.0 Design Basis Program 4.0 List of Actions 1.0 Nuclear Energy Institute (NEI) 96-05 Initiative In response to NRC Information Notice 96-17, "Reactor Operation Inconsistent with the Updated Final Safety Analysis Report," dated March 18, 1996, NEI 96-05, "Guidelines for Assessing Programs for Maintaining the Licensing Basis," was issued.

St. Lucie is currently proceeding with the review effort proposed by NEI, which will address progratnmatic and non-programmatic FSAR changes and assess the accuracy of two safety and two non-safety, risk significant systems described in the FSAR.

2.0 FSAR Reviews and Self-Assessments The following St. Lucie programs and initiatives are currently in progress or are completed.

2.1 FSAR Procedural Consistency Review During 1995 and 1996, NRC inspections found that plant operators were using procedures which had process steps not found in the FSAR and that the normal process steps found in the FSAR were not being followed. As a result of this event and other examples of similar procedural inconsistencies, FPL initiated an FSAR review for procedural consistency. This 1996 effort involved a comprehensive review of the FSARs (~ 8000 man-hours) with the primary focus being the identification of procedural processes in the FSAR and a comparison with the plant procedures to assure consistency.

The conclusion reached from this effort is that there were no operability issues which represented a substantial safety hazard. The final results of some condition reports are still pending, however, they have had operability reviews completed to ensure that they do not pose any unresolved safety concerns. The small number of CR-related inconsistencies identified indicate that there does not appear to be major discrepancies between the FSAR and the plant operating procedures.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 86 Instrument Setpoint Review Between 1991 and 1993, FPL performed an instrument setpoint review for the St. Lucie Plant to verify and document current setpoints. Plant operating and maintenance procedures were compared to the current field conditions and design documents to verify consistency. This effort is complete. The findings were formally addressed by the corrective action program and were closed.

2.3 Annunciator Review A plant annunciator summary review is currently in progress. This program willreview selected annunciators for the following alarm panels: control room, waste management, diesel generators A and B, steam generator blowdown building, boric acid concentrators A and B, waste concentrators, water treatment plant, the oxygen analyzer and liquid waste. This project is approximately 10% complete with over 200 annunciators reviewed. Discrepancies found in this review will be entered into the corrective action program by means of a condition report.

2.4 NRC A/E Inspection Findings FPL prepared for the NRC A/E Inspection effort by self-assessing three systems. There were several findings noted which were entered into the St. Lucie corrective action program. At the time of this response, the written inspection report has not been issued, but the NRC results were provided at a public exit meeting on January 28, 1997. The NRC exit results indicated that there were no major program weaknesses or failures, however, it was noted that future self-assessments would be beneficial.

3.0 Design Basis Program A design basis program was completed at St. Lucie on a selected number of systems. This program developed summary level documents which consolidated information contained in reference design bases sources (i,ecalculations, analyses, etc.), thereby making it more accessible in responding to plant operational needs. Design basis improvement activities will continue to be performed on an as-needed basis throughout the life of the plant. Such improvement activities have been primarily performed during the development of design modification packages and as corrective actions from various self-assessments and inspections through the years, and are expected to continue, although to a lesser extent, in the future.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 87 4.0 Lists of Actions St. Lucie has several planned and on-going actions which willprovide additional assurances that the units are operated and maintained within the design basis. These actions are:

4.1 The NEI Initiative (96-05) report will be issued to NEI by April 15, 1997.

4.2 The 1996 FSAR consistency review findings willbe resolved within a two year period.

4.3 The plant annunciator summary review is currently scheduled for completion in December 1998.

4.4 The recent NRC A/E Inspection findings will be addressed following issuance of the final inspection report and the corrective actions willbe provided by a separate letter or initiative as appropriate.

St. Lucie Ujtits I and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 88 Appendix A: Partial List of Acronyms A/E Architect/Engineer NRC Nuclear Regulatory Commission AFAS AuxiliaryFeedwater Actuation System 08'cMR Operations and Maintenance Reminder AFW Auxiliary Feedwatcr OE Operating Experience Report CM Configuration Management OEF Operating Experience Feedback CNRB Company Nuclear Review Board OSTI Operational Safety Team Inspection CR Condition Report PC/M Plant Change/Modification CRN Change Request Notice PCR Procedure Change Request CIRAC Commitmcnt Tracking PEG Production Engineering Group DBD Design Basis Document PEP Performance Enhancement Program DBRS Design Basis Reference System PM Preventative Maintenance DCR Drawing Change Rcqucst PMAI Plant Manager Action Item DER NRC Daily Event Report PMT Post-Maintenance Testing DWA Delivery Work Authorization POD Plant Operating Drawings EDG Emergency Diesel Generator PR Procurement Requisition EDS Electrical Distribution System PRA Probabilistic Risk Assessment EDSFI Electrical Distribution System Functional PRB Plant Review Board Inspection PSL St. Lucie Plant Units I and 2 ENG QI Engineering Quality Instruction PTN Turkey Point Units 3 and 4 EOP Emergency Operating Procedure PWO Plant Work Order EP Engineering Package QA Qua! ity Assurance EQ Equipment Qualification QI Quality Instruction ERT Event Response Team RCS Reactor Coolant System ESP Engineering Support Personnel REA Request for Engineering Assistance FCP FSAR Change Package RTST Real Time Support Team FPL Florida Power and Light Company SALP Systematic Assessment Of Licensee FRG Facility Review Group Pcifoflilailce FSAR Updated Final Safety Analysis Report SAR Safety Analysis Report GL NRC Generic Letters SASE Stand Alone Safety Evaluations HPSI High Pressure Safety Injection SATS System Acceptance Allover Status IA Instrument Air SDN Supplier Deviation Notice ICW Intake Cooling Water SEN SigniTicant Event Notification IEE Item Equivalency Evaluation SER Safety Evaluation Report In House Event SIT Safety Injection Tank IN NRC Information Notices SO Significant by Others Report INPO Institute Of Nuclear Power Operations SOER Significant Operating Experience Report ISI Inservice Inspection SQAD Special QA Document ITOP Implementor Turnover Package SSFI Safety System Functional Inspection J/LL Jumper and Lifted Lead SSC System, Structure and Component LER Licensee Event Reports SSH Significant Safety Hazard LPSI Low Pressure Safety Injection STAR St. Lucie Action Reports LOCA Loss of Collant Accident SWSOPI Service Water System Operational MEP Minor Engineering Package Performance Inspection MOV Motorpcratcd Valve TEDB Total Equipmcnt Data Base MPFF Maintenance Prcventablc Functional TQAR Topical Quality Assurance Rcport Failure TSA Temporary System Alteration MWt MegaWatts - Thermal NCR Non-Conformance Report NEI Nuclear Energy Institute NNS Non-Nuclear Safety Non-PC/M Non-Plant Change/ModiTication NPS Nuclear Plant Supervisor

St. Lucie Uttits 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 89 Appendix B Table B 4.7.1- QA Audit Findings AUDIT RESULTS QAS-JPN-91-8, "Juno Summatg Nuclear Engineering- The audit team focused on the St. Lucie Production Engineering Group (PEG), and the Production Engineering Juno Nuclear Engineering group. Auditors reviewedmodiflication packages, Group (PEG) for St. Lucie." associated safety evaluations, design bases documents, the environmental qualification program, and engineering procedures and quality instructions. Three audit Findings identified programinatic and administrative areas for improvement.

Plant changes and modiflication of operating plants.

Control of FPL originated design.

IdcntiTication of safety-related structures, systems and components.

Environmental qualification of electrical equipment Control of computer software Endmgr.

¹1. Some internal design interfaces with new JPN organizations are not documented.

¹2. Some design basis information in thc Total Equipmcnt Data Base gK)B) is not up-to~e.

¹3. Delivery Work Authorizations (DWAs) were issued for safety-related services to a supplier who was not on the approved supplier list for the location designated on the DWA.

The findings were formally addressed by the corrective action process and arc closed.

QAS-JPN-92-3, "Nuclear Summary; Engineering." The audit team focused on ASME Section XI, safety evaluations, design inputs, design bases, design analyses, engineering packages, special processes, engineering procedures, and the Juno Nuclear Engineering Quality Assurance Program. Three audit Findings were issued, but the deficiencies did not have a significant impact on safety or quality related activities.

Endings: ASME Section XI Safety classification and safety evaluations Technical specifications Design inputs/design basis Design analysis Minor engineering packages

¹1. Implementing proccdurcs and associated training did not always meet QA Manual requirements for the QA Records / Document Control Centers.

¹2. Non<estructive testing procedures conflicted with the Quality Instruction.

¹3. Receipt acknowledgments that were not received within 30 days of distributing QA records were not always followed up.

Thc findings were formally addressed by the corrective action process and are closed.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 90 Appendix B Table B 4.7.1- QA Audit Findings AUDIT RESULTS QAS-JPN-93-3, "Nuclear Sualrrtarg Engineering - PSL Design The audit team evaluated 12 engineering packages and their associated calculations, Control." along with documents and procedures associated with the PSL Probabilistic Risk Assessment (PRA). One Finding was issued for failing to update design documents following changes to the plant.

Engineering packages Mnor engineering packages Design and safety analyses Calculations Computer software control PC software control Endingr.

¹L Hardware changes were made to valves in the plant, but the Total Equipment Data Base gKK) was not updated to reflect those specific changes.

The findings werc formally addressed by the corrective action process and are closed.

QAS-JPN-95-1, "Nuclear Btttrrtrrarg Engineering Audit." This functional area audit evaluated engineering packages developed for the PSL Unit 2 cycle 9 refueling outage, real-time support for corrective action rcqucsts involving engineering, and long term enhancement projects such as the PSL 24 month fuel cycle.

Allmajor engineering processes and proccdurcs used to provide design output were evaluated. There were no audit findings or concerns identified during the audit.

Design control Engineering packages Minor engineering packages Calculations Design input verifications Environmental qualifications Engineering evaluations 10 CFR 5059 screening/evaluations 10 CFR 21 SSH evaluation and reporting Operability determinations safety classifications Vendor technical manual controls FSAR updating Design basis document updating TEDB Computer software control Eindirtgz None

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 91 Appendix B Table B 4.7.1- QA Audit Findings AUDIT RESULTS QSL-PM-9M6, April/May Snmmary; 1996 PSL QA Performance This routine audit focused on performance-based observations in the areas of Monitoring Audit. operations, maintenance, services and engineering, and resulted in 6 Findings. In the engineering area, discrepancies between the FSAR and PSL implementing procedures were identified, and procedures did not adequately address design requirements.

Endittgr.

¹1. The FSAR was not updated to reflect changes in radwaste storage practices.

¹4. No procedural guidance existed which described what constituted a loss of station air, and the required contingency actions to isolate an open penetration is station air was lost during core alterations.

The findings were formally addressed by the corrective action process and are closed.

QSL-PM-9648, May/June Snmnnuz:

1996 PSL QA Perfonnance This routine audit focused on performance-based observations in the areas of Monitoring Audit. operations, maintenance, services and engineering, and resulted in 3 Findings. In the engineering area, discrepancies between thc FSAR and PSL implementing procedures were identified.

Ending

¹1. The procedure and methods used to transfer resin from the blowdown building did not match the system operating description in the FSAR.

'Ihe findings were formally addressed by the corrective action process and are closed.

QSL-PCM-96-11, Audit of Sttmmarg the PSL program for review TYis audit focused on St Lucie Unit 1 Cycle 14 refueling activities. The "Reload" and implementation of plant modification package was reviewed, and QA identified that prercquisites for spent fuel change/modification pool cooling were not met prior to fuel movement. Additional modiflication packages packages, and stand-alone to be implemented during the rcfucling outage were reviewed, as well as the engineering safety modification package and safety evaluation review processes. QA found that the pre-evaluations. implementation review of design changes and safety evaluations by cognizant organizations is poorly defined, and do not provide adequate assurance that the requirements of these engineering documents will be captured and satisfied prior to implementation, resulting in 3 significant Findings.

Ending r.

¹1. Plant procedures and instructions do not adequately define the process for the review of PC/M packages by applicable departments, and the resolution of comments, during the PC/M preparation stage (before issuance to the plant).

¹2. Plant procedures and instructions do not adequately define the process for the review of PC/M packages by applicable departments, the resolution of comments, or the identification of plant procedures impacted by PC/M packages during the final review stage (after issuance to the plant).

¹3. Plant procedures and instructions do not adequately address the review of engineering safety evaluations for impact on plant procedures and instructions. No procedure exists which spcciTies how PSL reviews and implements requirements stated in safety evaluations.

The findings were formally addressed by the corrective action process and are closed

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 92 Appendix B Table B 4.7.1- QA Audit Findings AUDIT RESULTS QSL-PM-96-18, September ~

Bmnmxr.

1996 PSL QA Performance This routine audit focused on perfonnance-based observations in the areas of Monitoring Audit. operations, maintenance, services and engineering, and resulted in 6 Findings. In the engineering area, two Findings were indicative of less than adequate implementation of 10 CFR 5059 requirements.

Ettditgr.

¹3. Changes to plant systems were made using the minor maintenance process without adequately addressing configuration control requirements, or the requirements of 10 CFR 50.59, "Changes, Tests and Expcrimcnts."

¹5. The process for converting plant procedures to Guidelines, and for revising Guidelines do not meet 10 CFR 5059 requirements.

The finding ¹3 was formally addressed by the corrective action process and is closed.

Finding ¹5 is still being resolved.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 93 Appendix B Table C 3.12 - QA Audits and Vertical Slice Reviews AUDIT RESULTS QSL4PS-88-595, Vertical Sexual Slice Audit of the Unit 2 The audit team focused on the ICW system design, procurement of parts, and system Intake Cooling Water System. operation and maintenance, to assure system operability. The ICW system and components were inspected to evaluate configuration control, and to determine whether they were capable of performing the safety functions required by their design bases. Plant modiTications were reviewed to determine ifany unreviewed safety questions were introduced. System walkdowns verified as-built configuration. Pre-operational testing and hydrostatic testing met thc objectives of the design bases. The audit resulted in 5 Findings, none which involved discrepancies bctwcen as-built conditions and the design bases.

System design Procurement In-service Inspection Testing

¹1. Drawings did not reflect as-built conditions.

¹2. ICW system procedures contained discrepancies.

¹3. Improper or missing component identification tags.

¹4. External corrosion was not considered, and evidence of damage and deterioration were identifie.

¹5. Nonconforming conditions were not always documented.

Thc findings were formally addressed by thc corrective action process and are closed.

St. Lucie Uttits 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 94 Appendix B Table C 3.12 - QA Audits and Vertical Slice Reviews AUDIT RESULTS QS~PS-88424, Unit 1 Smnmatr.

and 2 Emergency Diesel This audit focused on EDG system design, procurement of parts, and system operation Generators and Safety- and maintenance, to assure system operability. The EDG system and components Related Switchgear Systems were inspected to evaluate configuration control, and to determine whether they were capable of performing the safety functions required by their design bases. Plant modifications were reviewed to determine ifany unreviewed safety questions were introduced. System walkdowns veriTied as-built configuration. The audit resulted in 5 Findings, which were formally addressed through the corrective action program.

Design operability Configuration control of design parameters PC/M safety evaluations Testing for operability Corrective maintenance configuration changes Individual timing relays and load blocks Load sequencer block voltage drop Electrical cabling Stretch power effects Eindiomr.

¹1. EDG fuel oil transfer pumps were tested, but not included in the IST plan.

¹2. EDG drawings did not reflcct as-built conditions.

¹3. EDG procedures describing systems and components were inaccurate.

¹4. Numerous items were found not tagged, or tagged incorrectly.

¹5. 'Inhere was no documentation available that demonstrated that the 16,450 gallon storage tanks for the Unit 1 EDGs met the design basis requirement of having sufficient fuel on hand to operate both EDGs in tandem for 4 days, or one EDG for 8 days.

The findings were formally addressed by the corrective action process and are closed.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 95 Appendix B Table C 3.12 - QA Audits and Vertical Slice Reviews AUDIT RESULTS QSLPS-89472, Summary:

Verification of the Unit 2 This audit focused on Reactor Coolant System (RCS) components and procedures, Reactor Coolant System with particular emphasis on instrumentation, evaluating 93 RCS instruments during Design, Operation and thc audit. Safety evaluation for system modifications were reviewed, and plant Maintenance, to Assure drawing were compared with as-built conditions. System walkdowns were performed, System Operability. and operating, maintenance and test procedures were reviewed. One Finding was issued regarding instrumentation.

Design operability Configuration control of design parameters PC/M safety evaluations Replacement pans Off-normal procedures Operating procedures maintenance procedures Plant drawings Annunciator drawing details Component tagging Zhtdtttgr.

¹1. One safety-related instrument and 8 instruments used to verify compliance with technical specifications werc not on a periodic calibration schedule.

The findings were formally addressed by the corrective action process and are closed.

QS LIPS-90-739, Bttmmarg Verification of the Unit 1 and This audit focused on the design bases of the AFW system as described in the FSAR, 2 AuxiliaryFeedwatcr System and configuration control of design information. Safety evaluations for system Design, Operation and modifications were evaluated, and the as-built configuration of the AFW Actuation Maintenance, to Assure System (AFAS) was inspected. Walkdowns of the Unit 1 and 2 AFW system were System Operability. performed, and Operating procedures and maintenance activities werc also evaluated.

One Finding documented inadequate procedures at Unit 2.

Design operability Configuration control of design parameters PC/M safety evaluations Replacement parts Plant drawings Annunciator drawings Component tagging AFW line-up procedure Entutgr.

¹1. AFW operating procedure position description for instrument root isolation valves did not match the drawings and as-found conditions.

The findings were formally addrcsscd by the corrective action process and are closed.

St. Lucie Upits 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 96 Appendix B Table C 3.12 - QA Audits and Vertical Slice Reviews AUDIT RESULTS QSL4PS-91-800, Suuuxxr.

Supplement 1, Vertical Slice This audit focused on the design, operation and maintenance of the Unit 1 and 2 Audit of the Unit 1 and 2 feedwater and main steam systems. The design bases, as described in the FSAR, were Feedwater and Main Steam reviewed to assure that configuration control was maintained. Safety evaluations for Systems. system modifications were reviewed to ensure unrcviewed safety questions did not exist. System walkdowns verified the as-built configuration, component identification, orientation, accessability, and general plant conditions. Operating procedures, plant drawings, maintenance activities and post maintenance testing were also reviewed.

Discrepancies between plant drawings and procedures resulted in 2 Findings.

Design operability Configuration control of design parameters PC/M safety evaluations Maintenance Testing Hadmu;

¹1. Flow diagrams did not reflect as-found conditions.

¹2. Plant procedures did not accurately reflec the plant configuration.

The findings were formally addressed by the corrective action process and are closed.

QSLNPS-91-821, Smltllarg Verification of the Unit 1 This audit focused on modiflication s to the RPS to ensure that system modifications Reactor Protection System did not introduce unreviewed safety questions. RPS set points were evaluated to ModiTications, Testing, ensure they were adequate to mitigate postulated accidents. RPS instrument Maintenance, and tolerances were evaluated. Maintenance and operating procedures were reviewed.

Procurement, to Assure System Operability.

RPS Set points PC/M safety evaluations Replacement parts Functional testing Instrument calibrations ZbldillgS. None .

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 97 Appendix B Table C 3.12 - QA Audits and Vertical Slice Reviews AUDlT RESULTS ISEG-PSL-A93417, ISEG Stannary Assessment Activities ISEG performed an independent risk assessment of nuclear safety during the 1993 Associated With the Spring Unit 1 refueling outage. Reviews were performed to ensure that the key safety 1993 St. Lucie Unit 1 functions for shutdown conditions (i.e. RCS inventory, decay heat removal, power Refueling Outage. availability, reactivity control, and containment integrity) were not compromised by outage work. One concern was identified involving the lack of verification of the design bases for the spent fuel pool as stated in the UFSAR, which could be jeopardized by outage work.

Canuhmhns; Outage scheduling and review efforts are performed with high regard to nuclear safety.

¹1 ~ Incorporate the spent fuel pool operability criteria into applicable outage management procedures.

ISEG-PSI A93419, ISEG Bumnturg Assessmcnt Activities ISEG performed reviews of the implementation of shutdown margin verifications, and Associated With the Spring pressurizer safety relief valve modifications.

1993 St. Lucie Unit 1 Refueling Outage. Cnnuhtsinnr.

Shutdown margin verifications were performed in accordance with technical specifications.

Pressurizer safety relief valve tailpipe loading modiTications resulted in zero leakage of the relief valves.

None ISEG-PSL-A93420, ISEG Summzg Activities Associated With ISEG provided independent oversight of the AC and DC power availability during the the Spring 1993 St. Lucic-1 outage, ensuring applicable requirements were satisfied. The integrated safcguards Refueling Outage. test was observed, and a plant modifications of the station blackout cross tie breakers was evaluated.

Canubtsinns; Power was adequately maintained during the refueling outage.

Fans for the pressurizer heater bus should be considered for routine cleaning each outage.

Prewperational testing of the Station Blackout plant modifications revealed problems that should have been dctectcd through the modiTication review process or the post-modiTication inspcctionhest process.

¹L Review the adequacy of thc process for ensuring that main circuit breakers are racked out and tagged whenever the 480- volt busses are cross-tied.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 98 Appendix B Table C 3.12 - QA Audits and Vertical Slice Reviews AUDIT RESULTS ITR 95402, Review of LER Sttmmarz:

335/94406: Containment This tcchnical review evaluated the causal factors and corrective actions associated Integrity Outside of FSAR with LER 335/94406. 'Ihe root cause of the event involved a design error made Assumptions Under Limited during the 1978 addition of the iodine removal system for Unit 1.

Circumstances due to Design Error. Cezbtsianr.

Causal factors for the loss of containment sump inventory in excess of the design basis were extensively developed.

Generic implications were adequately addressed.

Corrective actions bounded thc event.

None 1TR 96413, Safety System Summary; Walkdown (Partial): The QA Independent Technical Review group pcrformcd walkdowns of the Unit 1 Auxiliary Feedwatcr System, AFW system to verify that plant drawings reflecte as-built conditions.

Unit 1.

Condmhms:

list Quarter 1996 ESF system - The as-installed configuration was accurately rcflccted in plant drawings.

walkd owns] General cleanliness and material conditions were satisfactory.

None ITR 96415, Safety System Sumac.

Walkdown (Partial): Sections The QA Independent Technical Review group performed walkdowns of portions of the of Emergency Core Cooling Unit 1 high and low pressure safety injection and containment spray systems, to verify Subsystems in the Unit 1 that plant drawings reflected as-built conditions. The rcplaccment of the pressurizer Reactor Containment code safety valves under a modiTication package was also evaluated. Concerns were Building, and Pressurizer identifie regarding the plant's evaluation of an NRC information notice (IN), and 3 Code Safety Valve recommendations were made to address administrative problems.

ModiTications.

Cant:lusim:

[3rd Quarter 1996 ESF The as-installed configuration was accurately rcflccted in plant drawings.

system walkdowns] General cleanliness and material conditions were satisfactory.

Operating procedures properly implemented thc requirements of the UFSAR for the systems evaluated.

The PSL engineering evaluation of NRC IN 96-31 (regarding the potential for operation in an unanalyzed condition by operating with Safety Injection Tanks (SIT) cross-tied) did not contain a quantitative evaluation for the peak centerline fuel cladding tcmpcratures resulting from cross-tying Safety Injection Tanks (SITs). Condition Report CR 96-1726 was issued to track thc response.

¹l. Revise the pressurizer design basis documents to reflect the recent plant modification.

¹2. Enhance the minor engineering package preparation form.

¹3. Update the UFSAR to include the Unit 1 Cycle 13 large break loss of coolant accident analysis data.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-97-28 Enclosure Page 99 Appendix B Table C 3.12 - QA Audits and Vertical Slice Reviews AUDIT RESULTS PSL Nuclear Assurance Smnmarg Quality Report QR 964006, The QA Independent Technical Review group performed walkdowns of portions of the "Fourth Quarter 1996 ESF Unit 2 high pressure safety injection systems, to verify that as-built conditions were Walkdown of the Unit 2 Kgh reflected in plant drawings and met UFSAR requirements. Valve lineups were Pressure Injection System." performed using operating procedures, and components were verified to match design documentation.

Canclttsiatm:

The as-installed configuration was accurately reflected in plant drawings.

Valves were in their required positions. Minor identification tag discrepancies were noted.

41 installed components were verified to match the PASSPORT descriptions.

A potential lack of separation between HPSI discharge isolation valves was evaluated and found to be satisfactory.

Ecsttltr.

Satisfactory QSL-PM-96-18, September Sttmmarg 1996 PSL QA Performance This routine audit focused on performance-based observations in the areas of Monitoring Audit. operations, maintenance, services and engineering, and rcsultcd in 6 Findings. In the engineering area, two Findings werc indicative of less than adequate implementation of 10 CFR 5059 requirements.

Ezimgr.

N3. Changes to plant systems were made using the minor maintenance process without adequately addressing configuration control requirements, or the requirements of 10 CFR 5059, "Changes, Tests and Experiments."

N5. The process for converting plant procedures to Guidelines, and for revising Guidelines do not meet 10 CFR 5059 requirements.

Finding N3 was formally addressed by the corrective action process and is closed.

Finding N5 is currently in the corrective action process.

i~

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