ML20134L877

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

ML20134L877
Person / Time
Site:	Prairie Island
Issue date:	02/11/1997
From:	Wadley M NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9702190328
Download: ML20134L877 (60)

Text

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W Northem States Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401 1927 Telephone (612) 330-5500 4

th w ucr11,1997 10 CFR Part 50 Section 50.54(f)

U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60  ;

Response to Request for Information Pursuant to 10 CFR 50.54(f) l Reaardina Adeauacy and Availability of Desian Basis Information By letter dated October 9,1996, entitled " Request for Information Pursuant to 10 CFR 50.54(f) Regarding Adequacy and Availability of Design Basis Information," the U.S.

Nuclear Regulatory Commission (NRC) requested information pursuant to 10 CFR Part 50, Section 50.54(f) concerning facilities operated by the Northern States Power Company (NSP) and licensed by the NRC. The October 9th letter requests information concerning the processes applied at NSP's nuclear plants to operate and maintain the plants within their design bases and that reconcile deviations in a timely manner.

Exhibit A of this submittal provides such information for the Prairie Island Nuclear Generating Plant.

In addition, the October 9th NRC letter requested information concerning design review or reconstitution programs. Exhibit B of this submittal provides the requested information concerning the Prairie Island Configuration Management Program.

This letter contains no new NRC commitments, nor does it modify any prior commitments. The information provided in the exhibits to this letter is intended to describe processes and procedures as they exist as of the date of this letter. It is not intended to preclude subsequent changes following normal practices, or to require jVp NRC notification or consents for such changes other than those currently required. h 9702190328 970211 \

PDR ADOCK 05000282 I P PDR

i USNRC NORTHERN STATES POWER COMPANY February 11,1997 Page 2 Please contact us if you require further informatica. ]

C Michael D Wadley l' Vice President Nuclear Generation c: Director, Office of Nuclear Reactor Regulation Regional Administrator - Region ill, NRC Senior Resident inspector, NRC '

NRR Project Manager, NRC i J E Silberg State of Minnesota, Attn: Kris Sanda Attachments: Affidavit to the US Nuclear Regulatory Commission i

Exhibit (A) Prairie Island Nuclear Generating Plant Response to Required Information items (a) Through (e)

Exhibit (B) Description of Prairie Island Nuclear Generating Plant Design Basis Review Program i

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UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NOS. 50-282 and 50-306 RESPONSE TO REQUEST FOR INFORMATION PURSUANT TO 10 CFR 50.54(f)

REGARDING ADEOUACY AND AVAILABILITY OF DESIGN BASIS INFORMATION Northern States Power Company, a Minnesota corporation, by letter dated February 11,1997 provides its response to U.S. Nuclear Regulatory Commission (NRC) letter dated October 9,1996, with subject " Request for information Pursuant to 10 CFR 50.54(f) Regarding Adequacy E.r.d Availability of Design Basis Information." This letter contains no restricted or other defense information.

NORTHERN STATES POWER COMPANY By C ## h I Michael D Wadley Vice President Nuclear Generation On this / / ML day of febeuv i997 before me a notary public in and for said County, personally appe6 red Michael D Wadley, Vice President Nuclear Generation, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, and that to the best of his knowledge, information, and belief the statements made in it are true.

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Notary Public - Minnesota Sherburne County My Commission Expires January 31,2000

EXHIBIT A February 11,1997 Page1 Prairie Island Nuclear Generating Plant Response to Required Information items (a) Through (e)

CONTENTS Background 4 1.0 RequestforInformation 4 2.0 Permit and License History 4 3.0 Definition of Design Bases from NRC Information Request, footnote 4 4 4.0 Quality Assurance and Testing 5 l 5.0 Programs and Procedures 5 Part A 6 1.0 Summary 6 2.0 Control of Design Changes 7 3.0 Plant Alterations, Temporary Modifications and Bypasses 8 3.1 Alterations 8 32 Temporary Modifications 9 3.3 Bypass Control 9 4.0 Configuration Management 10 4.1 Modifications 10 42 Operating Procedures 12 4.3 Maintenance and Post Maintenance Testing Activities 13 4.4 Procurement of Spare Parts and Maintenance Services 14 5.0 Safety Evaluations 14 5.1 Design Change Safety Evaluations 14 5.2 Non-Modification Safety Evaluations 15 6.0 Verification of Program implementation 15 7.0 Site Engineering Staff 16 Part B 18 1.0 Startup Support for Prairie Island and initial Operation 18 2.0 Commercial Operation 19 2.1 Technical Competence of Engineering personnel 19 22 Surveillance Program 19 2.3 Preventative Maintenance Program 19 2.4 Post Maintenance Testing 20

. 3.0 Prairie Island Initiated Program Evaluations 20 3.1 Design Basis Document Development 20 32 Setpoint Study 21 3.3 Surveillance Testing Reviews 21 3.4 SWSOPI 22

a EXHIBIT A February 11,1997 Page 2 l 3.5 SSFV on Safety injection 22 3.6 ASME Section XI Program 22 4.0 QA Audits and Other Evaluations 23 i

5.0 NRC Evaluations 23 5.1 EDSFl 23 5.2 SALP Assessments 24 5.3 NRC Inspections 24 6.0 Operational Challenges to which Safeguard Functions Successfully Responded _ 24 6.1 SGTR - 1979 24 6.2 Loss of Offsite Power 25 Part C 26 1.0 Summary 26 )

2.0 Configuration Management / Design Basis Document 26 3.0 Self-Assessmentinitiatives 27 3.1 SWSOPl 27 3.2 SSFV on Safety injection 28 3.3 Station Blackout 29 3.4 Safety Evaluations 29 3.5 Plant Engineering Self Assessment 30 4.0 Walkdowns 30 4.1 DBD Walkdown Results 30 4.2 SWSOPl 31 4.3 IE Bulletin 79-14 31 4.4 The Electrical and l&C Drawing Upgrade Project 32 5.0 Audits 32 6.0 External Assessments 33 6.1 EDSFl , 33 6.2 NRC Inspection of Moiification Activities 34 6.3 NRC Inspection of Sciety F <aluations 35 6.4 NRC Inspections of DBD Activities 35 7.0 SALP 36 Part D 37 1.0 Summary 37 2.0 Review of Operational Events 37 2.1 Operating Experience Assessment (OEA) 37 2.2 Engineering issues 37 3.0 Self Assessment Problem identification Sources 38 3.1 Employee Observation Reports (EOR) 38 3.2 Follow-on items (FOI) 38 3.3 Plant inspection Program 38 4.0 Audit Activity 39 4.1 Intemal Audit Program 39 5.0 Corrective Action Programs 39

EXHIBIT A

February 11,1997 Page 3 5.1 Nonconfonnance (NCR) 39 5.2 Root Cause Evaluations 40

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6.0 Processes for Reporting to the NRC 40

, 6.1 Reporting 40 Part E 41 1.0 Prairie Island Response 41 2.0 Prairie Island Design and Configuration Management Controls 41 3,0 Prairie Island Self Assessments 42 3 4.0 Quality Assurance Assessments 43 5.0 NRC Inspections 44 I 6.0 Prairie Island Current initiatives 44 6.1 USAR Review Project . _44 j 62 Safety Evaluation Program Enhancements . _45

7.0 Conclusion , 45 l References 47 Acronyms 50 l

EXHIBIT A i February 11,1997  !

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Background

1.0 Request forinformation By letter dated October 9,1996, entitled " Request for Information Pursuant to 10 CFR 50.54(f) Regarding Adequacy and Availability of Design Basis Information," the U.S.

Nuclear Regulatory Commission (NRC) requested information pursuant to 10 CFR Part 50, Section 50.54(f) concerning facilities operated by the Northern States Power Company (NSP) and licensed by the NRC. The October 9th letter requests information concerning the processes applied at NSP's nuclear plants to operate and maintain them within the design bases and to reconcile deviations in a timely manner.

2.0 Permit and License History By application dated March 29,1967 to the Atomic Energy Commission (AEC), NSP requested a permit for construction of the Prairie Island Nuclear Generating Plant (Prairie Island). The AEC issued construction permit CPPR-45 and CPPR-46 to NSP l on June 25,1968. By Amendment 7 to the March 29,1967 application, NSP requested all necessary AEC licenses to operate Prairie Island. The Safety Evaluation Report for Prairie Island was issued by the AEC Directorate of Licensing on September 28,1972. The Prairie Island interim operating license, License DPR-42, was issued on August 9,1973. Facility Operating License DPR-42 for Unit I was issued April 5,1974.

Facility Operating License DPR-60 for Unit 2 was issued October 29,1974.

The Prairie Island licenses were issued based on the design bases as documented in the Prairie Island Final Safety Analysis Report (FSAR) submitted with Amendment 7 to the application and supplemental information as submitted in amendments 8 through

38. The initial licensing of Prairie Island predated much of the current regulatory i framework concerning design bases documentation. The Safety Evaluation for dockets 1 50-282 and 50-306 issued by the Directorate of Licensing, U.S. Atomic Energy Commission finds the plant "in accord with the Commission's General Design Criteria, l Quality Assurance Criteria, Safety Guides and appropriate industrial codes and standards, and that any departures from these criteria, codes, and standards have been identified and justified."

3.0 Definition of Design Bases from NRC Information Request, footnote 4

" Design bases mean that information which identifi9s the specific functions to be performed by a structure, system, or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design..."

The design bases of a facility, as so defined, is a subset of the plant licensing bases and is contained in the FSAR. Information developed to implement and support the design bases is contained in other documents, some of which are docketed and some of which are retained by the licensee.

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EXHIBIT A Febmary 11,1997 i;

Page 5 i 4.0 Quality Assurance and Testing

! The Prairie Island design bases documented in the license application was

{ implemented during plant construction via system and equipment specifications, which

were translated into construction drawings. Reasonable assurance of plant j construction in accordance with the design basis was provided via NSP Quality Assurance audits and selected contractor QC inspections of construction activities and i by the performance of pre-operational and startup testing. The plant's Technical Specifications, issued as Appendix A and B to the Prairie Island Facility Operating

l License, provided further reasonable assurance that the plant would be operated within

! significant parameters consistent with the design basis.

Additional means to maintain the plant's design bases was provided by NSP's compliance to the requirements of 10 CFR 50, Appendix B. This was expanded with Revision 2 of the Operational Quality Assurance Plan (OQAP), effective November 15,

1977, which conditionally invoked ANSI N18.7, and the N45.2 series standards,

including N45.2.11, as modified by pertinent NRC Regulatory Guides and NSP

compliance positions as specified in the OQAP. The codification of 10 CFR 50.71e

further promoted maintenance of the design bases.

ld 5.0 Programs and Procedures

' The NSP Quality Assurance program described in the OQAP is implemented by NSP through Corporate Nuclear Administrative Control Directives (N1 ACDs) and at the

Prairie Island site through plant Administrative Work Instructions (SAWis) and plant j procedures. These controls provide reasonable assurance that Prairie Island is

maintained and operated consistent with the plant's design bases. In addition, the NSP

! Quality Assurance program provides reasonable assurance that discrepancies in operation or configuration with respect to the plant's design bases are identified, 3 evaluated and corrected in a timely manner. Additional discussion conceming the i attributes of the Prairie Island programs and processes which address the information y requested in items (a) through (e) of the Commission's October 9,1996 letter pursuant j to 10 CFR 50.54(f) is provided as follows.

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EXHIBIT A February 11,1997 Page 6 Part A Description of engineering design and con &guration controlprocesses, including those thatimplement 10 CFR 50.59,10 CFR 50.71 (e), and Appendix B to 10 CFR Part 50 1.0 Summary NSP has instituted design and configuration management controls as part of its Operational Quality Assurance Plan [OQAP). These controls provide reasonable assurance that Prairie Island's systems, structures and components, and their modification, maintenance, surveillance and operation, implement the plant's established design bases. They include provisions for evaluating, reporting and processing changes to and discrepancies with the SAR' requirements, in accordance with 10CFR50.59 and 10CFR50.71(a).

The principal QA program processes at Prairie Island for applying Appendix B and the invoked ANSI N45.2 standards are the plant's Administrative Work Instructions

[5AWis), as governed by the corporate Administrative Control Directives [N1 ACDs).

These are supplemented by departmental work instructions and plant procedures.

Changes to the design of plant systems, structures or components [SSCs), are controlled by the plant's SAWis. Integrated in these instructions are steps addressing compliance with the design requirements including the plant's design bases. These include explicit checklist items in the design input and design verification processes.

Provisions in the plant's SAWis provide reasonable assurance that design documents, such as drawings and specifications, have had the benefit of the above process controls utilized in making changes to plant configuration. Further, the plant's 5AWis include steps that provide reasonable assurance that operating procedures, tests and setpoints reflect changes in plant configuration. Procurement related controls invoke appropriate design and configuration controls on Architect / Engineers [A/Es), vendors and contractors that perform work, and provide for monitoring of compliance.

Compliance of site activities is monitored by NSP's Generation Quality Services (GQS]

group, as specified in their procedures. The foundation of these program controls were introdtced in early plant ACDs, and have been built upon into the present plant 5A Wl's.

The p ant's 5AWis provide controls for processing changes to the design bases requirements. These 5AWis specify the development of Safety Evaluations and reviews by the Operations Committee for those changes, and items that pose a potential unreviewed safety question, as required by 10CFR50.59. Changes to the SAR are

' SAR as used in this document refers to the FSAR and USAR. Explicit document titles are given [e.g. USAR] l when it is important for clarity, such as in describing the history of SAR changes at Prairie Island. This discussion also applies to the Independent Spent Fuel Storage Installation Safety Analysis Report [ISFSI SAR], '

and to 10CFR72.48 reviews, where 10CFR50.59 reviews are discussed. l l

EXHIBIT A February 11,1997 i Page 7 formally processed and submitted to the NRC for concurrence, as required by 10CFR50.71(a). The SAWis that relate to design control and configuration management processes are included in the References listing, and are summarized below, i 2.0 Controlof Design Changes Controls on the processing of design changes' at Prairie Island are delineated in the SAWI 6. series instructions. Supplementary details for specific engineering functions, l such as development of design outputs and design verification, are defined in the Site l Engineering Manual. An overview of the design change process defined in the 5AWis

! follows.

1 A design change is initiated in accordance with the Integrated Planning Process, with the project scope defined in a Solution Team Report. As practical, the members of the Solution Team transition to the design change team to provide continuity during the design change process. Typically, one of the first activities of the design change l project is the identification of applicable design inputs, which are required to be l identified and documented throughout design change development. The Design input l Applicability Checklist is required to be completed to aid in the identification of the l design inputs, and is divided into specific topic areas. Several of these topic areas l invoke the SAR and Design Bases Documents, as potential input sources.

The design engineers are to utilize the identified design inputs in the development of l l

design output documents. Output documents include drawings, calculations, specifications, technical manuals, electrical schematics, cable data and system descriptions. Design output documents are required to be issued for review, and the Document Routing and Review Sheet is to be utilized, when deemed to be appropriate.

This form includes an overview checklist, one item of which is " licensing" requirements.

Safety Related and Augmented Quality

• design output documents are to receive a design verification, by a knowledgeable person satisfying prescribed independence criteria. Design verification can be performed by a design review, attemate calculations, or qualification testing. Most commonly, design reviews are utilized. The design reviewer is required to consider compliance with the Technical Specification, SAR, or pending SAR submittals. After resolution of the review and verification  !

comments, the design output is to be approved. In addition to review and verification, drawings, plans and specifications from which construction is to be accomplished ,

typically are required to be certified by a Minnesota licensed Professional Engineer; exceptions include output relating to "off-the-shelf" items. For design work performed 8 '

The terms " design change" and " modification" are interchangeable. Past definition differences between these two terms are no longer applicable.

( ' Augmented Quality applies to items or services that do not perform a sarety related function, but are subjected to special NSP or NRC regulatory imposed requirements. These include items or sersices that are non-safety QA relas4 fire protection related, security related,10CFR71 related, and 10CFR72 related.

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! February 11,1997 Page 8 by Architect / Engineers [A/Es], Prairie Island is required to invoke similar controls on their safety related activities. These requirements are to be invoked on A/Es through NSP's procurement process, with the adequacy of A/E compliance being monitored through NSP's Quality Services activities.

To provide real-time control of the design document changes during implementation of the design change, revisions to approved design output documents, such as drawings released for fabrication, construction or installation, are to be processed utilizing a Engineering Change Request [ECR]. The ECR process specifies review and approval of the change by the design organization. Prior to close-out of the ECR, confirmation of "as built" conformance to the approved design change is required. Implemented ECRs are tracked to provide reasonable assurance that they are reflected in the final design change outputs.

In order to inform plant organizations of the scope and impact of a developing design change and to provide for a project review, the project is required to develop and issue a Project Description / Safety Assessment [PD/SA). This document provides a concise definition of the scope of the design change, including civil, mechanical, electrical, nuclear, operational and l&C attributes, as appropriate. It indicates quality classifications, such as safety related, or fire protection related, and any boundaries on these classifications. It discusses the impact of the changes on plant systems, structures and components, including references to plant drawings. Adequate information is reqaired to support a safety review by the Operations Committee, if one I is determined to be required. The PD/SA also is required to provide the technical bases for the change, and possible failure modes to be considered in an accident .

review. In addition, it is to describe adequately the impact on the existing SAR '

(including pending SAR submittals), enabling a direct translation into a SAR change, if l required. i i

To provide reasonable assurance that the design change complies with plant administrative and design bases requirements, the plant's SAWis require that a Design Change Package be prepared and issued for review. The Design Change Package includes design change administrative forms, some of which were discussed above, and the PD/SA. Reviews of this package include a Document Control Review and a Design Bases Document review. The latter requires that the package be reviewed against applicable Design Bases Documents [DBDs), for reasonable assurance that the design change is in compliance with the current design bases. Instructions also specify that changes to facility design be reviewed to determine the need for a Safety Evaluation and review by the Operations Committee in accordance with 5AWI 3.3.2.

3.0 Plant Alterations, Temporary Afodifications and Bypasses 3.1 Alterations Alterations, as opposed to plant design changes, involve the replacement of a component with one that meets the same specification as, but is not identical to, the

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$ Page 9 l original component. This activity is of limited scope, and does not necessitate the

development of a Design Change Package. Updating of affected plant documents, i such as Technical Manuals, drawings, and Data Files is also required. Recent j changes in the SAWis have eliminated Alterations, however for those initiated prict to i

this change, these requirements are still in effect until they are completc d and closed out.

i 3.2 Temporary Modifications 1

Prairie Island's SAWis provide reasonable assurance that Temporary Modifications [T-Mods) do not compromise the plant's design bases or safe operation. T-Mods are required to be of minor scope and of short duration. T-Mods are processed using a Temporary Modification Request. T-Mods are required to be approved before installation and subsequent removal. As appropriate, T-Mods may require temporary

changes to plant procedures, drawings and the training program. T-Mods are required j to be clearly identified and reviewed periodically to ascertain continued need.

i The plant's 5AWis require design verification of the T-Mods to provide confidence that

modified equipment will operate as intended, and that plant and personnel safety will i

not be adversely affected. The plant's SAWis for T-Mods require that any changes to

the plant's Technical Specifications or SAR introduced by the modification be identified.

Reviews of T-Mods are required, to determine the need for a Safety Evaluation and

, review by the Operations Committee, in accordance with 5AWI 3.3.2. If T-Mod j installation or removal introduces a Technical Specification change or an unreviewed

safety question, approval by the NRC is required before implementation. T-Mods are to be implemented utilizing a plant Work Order. Requirements for processing changes to procedures, drawings, data files and licensing documents parallel those for

permanent design changes, including a future needs turnover checklist. '

3.3 Bypass Control As with design changes, instructions for the application of bypasses provide l reasonable assurance that design bases requirements are not compromised.

- Bypasses are required to be reviewed to confirm that they will leave the plant in a safe condition, and to determine whether a Safety Evaluation in accordance with 5AWI 3.3.2 l is required. Approval is required for the installation and removal of bypasses.

j independent verification is to be conducted where appropriate.

i The plant's 5AWis require confirmation that bypass placement or removal does not l compromise plant safety requirements or violate the Technical Specifications.

Bypasses are required to be physically tagged, and entered in the Bypass index. A l j review of the Bypass Index is required to be performed every three months, with the ,

results reported to the Operations Committee. The Operations Committee is required to periodically review the status of open bypasses.

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4.0 Configuration Management i

j 4.1 Modifications j 4.1.1 Design Change implementation l During implementation of a design change, consistency with the configuration represented in the Design Change Package is addressed through the dccument control i system. The plant's SAWis specify that only design output do::uments that have j received the required reviews and approvals are to be utilized on design changes to l systems, structures and components. Most changes to plant configuration are detailed

} in drawings. Verified and approved drawing revisions are to be entered into the

Controlled Drawing Files.

} Another element in design change configuration management, is the Work Order l Package. This package is the instructional and authorization package from which the j actual design changes are implemented in the plant. Work Order Package l requirements defined in the plant's SAWis stipulate that drawing references in the package "shall" be developed from controlled drawings. In addition, it stipulates that drawings not identified as design output or controlled by design change process l procedures, shall likewise be "be controlled, uniquely identified, reviewed and j approved ...," and attached to the Work Order Package.

I j Large and complex projects may racessitate the development of specific configuration j management controls. For examp,u, the recent Station Blackout / Electrical System i Upgrade [SBO/ESU) project, which added two new emergency diesel generators and

! upgraded the plant's 4KV safeguard buses, had a dedicated organization and set of i Project Procedures (PPs) and work instructions for directing design and configuration control activities. These procedures paralleled the design and configuration controls in ;

j place at the time, including use of ECR and controlled drawing files, with customized j j organizational and administrative provisions. Implementation was verified through a

dedicated QA evaluation program. In addition, the project had matrixes that delineated I regulatory requirements and commitments, and tracked their implementation in plant j configuration, system testing and operational procedures. Future large and complex l projects will be evaluated by Prairie Island management to determine if specific i configuration management controls would be appropriate.

i i 4.1.2 Modification Activities Performed by External Organizations

NSP's QA Program requires that organizations extemal to NSP, such as contractors, i vendors or A/Es, involved in Safety Related, and to a lessor extent Augmented Quality

! design changes, have QA programs in place covering their design and configuration j control activities, perform them under Prairie Island's program, or a combination

thereof. These programs are to be evaluated and monitored for adequacy in l accordance with NSP's QA Program.

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4 s EXHIBIT A February 11,1997 Page 11 i Specifications, which delineate technical and QA program requirements to be invoked l on external organizations, are design output documents and are reviewed and verified ,

j accordingly. Specifications are controlled by the plant's SAWis, which prescribe a formal index of the latest approved versions. Technical and QA requirements not

! included in specifications are to be delineated in procurement requisitions and

supplementary documents, which are to be formally reviewed, approved and translated  ;

j into contract documents.

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! As deemed appropriate, source verification of the implementation of the design change 4 requirements by external organizations is to be performed through NSP's Supplier QA

group. On special occasions, site QC inspectors may perform source inspections, if c requested by the modification team. Routinely, site QC inspectors are to perform identified receipt inspections of procured items. In addition, necessary post-receipt ,

! testing is to be identified on the requisitions, except for Commercial Grade items, and '

i performed as specified. These quality verification processes provide reasonable  ;

! assurance that design bases requirements identified in the Prairie Island Design j

[ Change Package are implemented by external organizations. l i 4.1.3 Post Modification Testing l l Design change testing is performed, to the extent appropriate, to demonstrate that a l

! modified SSC meets its design intent and will perform its design function. These tests

! include those for evaluating compliance with design change details, such as continuity i and hydro tests, and those confirming functionality, namely start-up and pre-operational t

tests. The latter confirm that the component will perform as required to support defined j system or structure functions. These tests are typically performed under conditions l representative of normal operating conditions, but they may also simulate abnormal or

! emergency operating conditions. Within practical limits, this tests provide reasonable

! assurance that the modified systems, structures and components will comply with the j design bases

! 4.1.4 Administrative Support Systems l Plant administrative systems aid in configuration management during the development i and implementation of design changes. Approved design analysis are listed in the

! Prairie Island Analysis Index data base, which is required to identify the analysis l number, revision, title, and affected system, structure and component.

j Component information that supports design change activities is maintained in the

Plant's Component Data Files on the site computer system. These data files include l technical descriptions and data on individual plant components, including their Q-List ,

l extension and any special concerns. The Q-List extension is the safety related i classification of a component and its individual parts, as derived from their individual

contribution to the safety related functions of the parent system er structure. Plant  ;

Component Data File updates are reviewed and controlled in accordance with the i j plant's SAWis. Plant QA records and historical technical data that support design i j change activity are maintained in accordance with the plant's record management 4

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February 11,1997 1 Page 12 procedures. Certain resource documents are required to be maintained up-to-date in hard copies; these include the SAR, Technical Specifications, DBDs, Technical l Manuals, plant procedures and SAWis.

4.1.5 Design Change Turnover and Close-out

The plant's SAWis provide controls to prevent a modified system, structure or

! component from being placed in operation until applicable supporting analysis, l documents, operational procedure changes, and related items are in place. These controls utilize two forms, the Tumover Checklist and Record Log. The Tumover Checklist documents the completion of relevant tasks, including the updating of equipment numbering and labeling, completion of needed training, completion of calculations and analysis, transitional drawing files updated to reflect the change, updating of operational procedures, and identification of required SAR changes.

Documents that have been developed or affected through a design change are itemized on a Record Log; these include Work Orders, Drawing Requests, Technical Manuals, Setpoints, SAR text, operations procedures, and Technical Specifications changes. Any items on tne Records Log required to be completed prior to system operation, such as operational procedures or Setpoints affected by the design change, must be so indicated. The Record Log is attached to the Tumover Checklist, and instructions require the completion of items identified as required for system operation, prior to tumover. Open items not required for tumover, but required for design change close-out [ future needs), are to be itemized on a punchlist attached to the checklist.

For design changes performed by non-plant organizations, a final plant review of the Tumover Checklist is required to confirm that the punchlisted open items will not affect the safe operation of the portion of the design change being turned over.

A plant design change may not be closed-out until open items identified on the punchlist attached to the Turnover Checklist are completed. This requirement provides added confidence that details in plant documentation affected by the design change are reconciled. To control this process, a Close-out Checklist is to be completed.

Checklist items include the updating of drawings, specifications, and technical manuals not required at tumover; the updating of the Safe Shutdown lists and EQ Master List, and that a SAR update, with a summary of the Safety Evaluation being sent to the Licensing and Management issues Department [ Licensing].

4.2 Operating Procedures 4.2.1 Control of Operation Procedures The plant's SAWis for generating or revising plant operating procedures provide confidence that they reflect changes in plant configuration and comply with the design bases. These include Emergency Operating Procedures, Operating Procedures, Alarm Response Guides, Maintenance procedures and Surveillance procedures. New procedures and procedure changes are required to be reviewed by a knowledgeable person other than the preparer. Procedures are to receive a technical review to

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! determine if they introduce a deviation from SAR designated steps," " system

, operation," or " commitments," thus necessitating a Safety Evaluation in accordance with 5AWI 3.3.2. The technical review also is required to confirm a procedure's

compliance to the Technical Specifications, technical accuracy and adequacy, j compliance with the plant's Quality Assurance requirements, usability or workability,  ;

and that the appropriate plant configuration is maintained,. In addition to this review, l l procedures fulfilling Technical Specification requirements are to be reviewed by the l Operations Committee for overall adequacy.

) As an additional control, the plant's SAWis specify that " applicable procedures are to be reviewed following any design change to a system, or unusual plant incident...". In i addition, operating procedures are required to be reviewed periodically by the

Operations Committee, in accordance with Technical Specifications. These reviews j include consideration of Technical Manuals and outstanding Temporary Memos, and j~ whether a Safety Evaluation is required, and if so, whether it has been appropriately completed.

i 4.2.2 Setpoint Changes

! Establishment and changes to setpoints are controlled to provide confidence that they

! conform to the design bases. Setpoints are identified in the computerized Calibration l Card /Setpoint File Program [CCSF). As discussed above, necessary changes to l 4 equipment setpoints resulting from a design change must be identified on the Record Log. Additions or changes to setpoints are to be initiated with a Setpoint Change i Request [SPCR] and a Work Order. The SPCR is required to be reviewed against the 4

requirements of the SAR and Technical Specifications. Instructions require that i setpoints be reviewed to verify compliance with 10CFR50.59 and 10CFR72.48, and ,

j approved. The Alarm Response Guide is to be updated, when applicable. When i {

deemed appropriate, a Safety Evaluation or Operations Committee review is to be '

l initiated.

] 4.3 Maintenance and Post Maintenance Testing Activities The plant's SAWis require that maintenance work performed on installed systems, structures or components be managed through the Work Control Process. This work is initiated through generation of a Work Order. The scope of the identified maintenance j work is to be reviewed to determine whether a design change is required. Critical Work, as defined by the plant's SAWis, is reviewed prior to implementation to provide confidence that Technical Specifications are addressed and that plant safety is not jeopardized. Work on Safety Related and Augmented Quality systems, structures and components is to be inspected by Quality Services, as deemed appropriate, to provide reasonable assurance that programmatic and Work Order requirements are implemented.

After completion of maintenance activities, appropriate testing is performed to confirm the ability of the respective system, structure or component to perform its intended function. This testing is typically performed under conditions representative of normal

I.

Fl HIBIT A February 11,1997 Page 14 operating conditions, but it may also simulate abnormal or emergency operating conditions. Maintenance controls and post maintenance testing provide a degree of '

reasonable assurance that maintained systems, structures and components will comply with the design bases, i

4.4 Procurement of Spare Parts and Maintenance Services Activities for obtaining maintenance services and spare parts are the same as discussed above for modification activities performed by extemal organizations. In addition, the plant's SAWis require spare parts to be purchased to a specification,  ;

usually the original specification. If the original part is not available, an evaluation is to l be performed by a qualified individual to assess the suitability of the replacement part i to perform the required function, thereby maintaining the design basis. Receipt Inspection by Quality Services for Safety Related and Augmented Quality j procurements provides reasonable assurance that the part or service will conform to '

the purchase specification.

5.0 SalietyEvaluations 5.1 Design Change Safety Evaluations Design change Safety Evaluations are to be prepared in accordance with SAWI 3.3.2.

This instruction also delineates the method for determining if activities affect the facility i as described in the SAR (or pending SAR submittal) and whether a Safety Evaluation is required.

When it is determined that a Safety Evaluation is appropriate, it is to be developed with the following contents:

. A description of the change, including impact on the plant while being 4 implemented, as well as after it has been turned over for operation.  ;

e A description of design considerations that mitigate failures or consequences, such as lockouts, alarms, material selection, or redundancy, '

administrative controls, QA requirements and previous experience with similar systems, structures or components.

e identification of the systems, structures or components affected by the change, and their function.

. Identification of the SAR, Technical Specifications, SER or other commitments to the NRC that apply to the change, and what design basis accidents were reviewed for impact.

. A statement as to whether there is an unreviewed safety question, with documented answers to the seven questions provided in the instructions.

. A statement as to whether the change involves a deviation from the Technical Specifications.

EXHIBIT A February 11,1997 Page 15 Safety Evaluations are required to be reviewed prior to approval by a knowledgeable individual other than the preparer, for adequacy, including SAR and Technical Specifications considerations, design bases requirements, Nuclear Analysis and j Design [NAD) analysis, and other special considerations.

The Safety Evaluation is forwarded with the Design Change Package to the Operations Committee [OC] for review. The OC reviews the design change Safety Evaluation, including possible inconsistencies with the SAR (or pending SAR submittals). The OC is to determine if the change is inconsistent with descriptions of plant configuration or procedures provided in the SAR, or introduces new tests or experiments, and whether it creates an unreviewed safety question. Subsequently, the Safety Evaluation is to be forwarded to the Safety Audit Committee [ SAC] for review. Those designated as a  !

potential unreviewed safety questions and changes to the Technical Specifications require SAC approval before implementation. The charter and membership of the SAC and procedures for its plant safety review are defined in Licensing procedures. j The completion and outcome of these safety reviews are to be indicated in the Design Change Package. If the reviews determine that the design change introduces an '

unreviewed safety question or necessitates a change to the Technical Specifications, I there are three options. The first is to cancel the design change, and the second is to I redesign it so that it no longer involves an unreviewed safety question. The third option is to submit the design change to the NRC for approval, which is required prior to implementation of the change.  !

5.2 Non-Modification Safety Evaluations Non-Modification Safety Evaluations [Non-Mod Safety Evaluation] are operating )

procedure changes, new tests or experiments and other items not associated with a '

design change. Such items feed into the Safety Evaluation process through various i means, such as through the Proceiure Submittal Form, a Nonconformance Report or  ;

Follow-on item [FOl]. A Non-Mod Saiaty Evaluation is processed utilizing a Safety Evaluation (Non-Modification) form. Requirements for reviewing, approving and  :

submitting them are the same as for design change Safety Evaluations.  !

As Non-Mod Safety Evaluations are not associated with a design change package, special configuration management controls are included for them in 5AWI 3.3.2. 1 Actions requiring completion as a result of the Non-Mod Safety Evaluations, such as l the revising of a drawing or plant procedure, and tests or training needing to be performed, are to be identified on a future needs checklist. Future needs items are to be tracked until closure. Non-Mod Safety Evaluations are to be controlled, indexed and maintained per plant SAWis.

6,0 Verification of Program implementation There are several programmatic activities that assess the implementation of the above controls, and the integrity of configuration changes. Principally these are design

l 4

EXHIBIT A i 1

February 11,1997 Page 16

! change walkdowns, close-out approvals of the Design Change Packages, QA Audits, .

j- and pre-operational tests. Walkdowns of installed design changes for compliance with '

j approved design documents has been a practice prescribed by NSP since original

construction. The plant site engineering presence has facilitated this practice.

l Walkdown considerations and sign-offs of completed work are prescribed by the plant's  ;

5AWis for design changes as part of Work Control Process. For work performed '

i under the cognizance of Nuclear Generation Services [NGS) and its predecessors, this t l' practice was primarily implemented by the contractors and monitored by NSP.

s l The close-out approval of the Design Change Package indicates that the " requirements

! of the design change have been met." In addition, the sign-off on the design change l Close-out Checklist signifies that documents that maintain facility configuration have l been updated, and a SAR update was submitted to Licensing, if required.

I

{ As the procedures discussed above are part of Prairie Island's QA program, their j implementation is the subject of QA auditing, by NSP's Generation Quality Services. A i summary of audit activities is provided in Part C. In addition to QA audits and j evaluations, site Quality Services includes a QC group that performs inspections as

! directed by the plant's SAWis. These include inspections by qualified, and as required, j certified personnel of modification and maintenance activities. inspection coverage of j these activities include the use of correct materials and parts, adequate completion of i documentation, and that the work is completed per Work Order requirements and

! applicable procedures. Inspection activities also involve quantitative verifications, and l NDE. As discussed above, GQS QC inspectors are also involved in source and receipt

. inspections of procured items, and in source inspections on special occasions, as I discussed earlier. This wide scope of audit and inspection activities provide i reasonable assurance that the design bases, as reflected in plant design and

{ maintenance documents, are being incorporated in plant systems, structures and l components.

! 1.astly, as discussed above, post modification testing, including pre-operational tests, l and post maintenance testing are prescribed by the plant's SAWis. As appropriate, the

acceptance criteria in these tests confirm compliance to performance and set-point j_ criteria that are important to the safe operation of the plant. These are often derived j from the design bases, and when deemed appropriate prescribed in license documents l submitted to the NRC. Similarly and as discussed above, plant instructions prescribe l post maintenance testing. Post modification and post maintenance tests confirm that

affected systems, structures and components will meet design requirements, including those derived from the plant's design bases.

, 7.0 Site Engineering Staff l An important feature at Prairie Island for promoting compliance to design bases

, requirements is the System Engineers. This organizational position has existed since

} initial plant operation. System Engineers have been the focal point for technical

} integration and regulatory compliance of activities associated with operation, i

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l EXHIBIT A j Febniary 11,1997

! Page 17

! maintenance, monitoring and design changes of plant systems. The majority of current i System Engineers hold company sponsored SRO Certification, and many in the past held Senior Reactor Operator Licenses.

1 The Prairie Island Plant Engineering Handbook states that, "the System Engineer has overall responsibility for the nuclear safety, reliability and availability of assigned systems," and is the " recognized expert on his or her systems." As the title implies, the System Engineer is assigned and dedicated to system [s] important to plant safety.  !

System Engineers are to be knowledgeable in system design requirements, including those in the SAR, and Technical Specification. Their defined responsibilities include l l " ensuring regulatory compliance," and " evaluating Technicral Specifications for l appropriateness and compliance."

System Engineers participate in plant meetings on operation, design change and l

l. licensing issues, assist persons developing Design Change and Work Control '

! Packages, and respond to NRC commitments and concerns, that involve or impact their l l systems. They provide technical assistance to other site groups, including the i

! Operations Committee, and are a key element in the plant's compliance with the design 1 bases.

I 1 1  :

in addition to System Engineers, NGS design engineers are also represented on site at l~

Prairie Island. These engineers are discipline oriented, and are knowledgeable in industry and plant technical requirements, including design bases. Many of these l l engineers were stationed at the Prairie Island during initial operation, and have been  ;

{ there continuously since the TMI initiative was implemented. The majority of current i

NGS engineers hold company sponsored SRO Certification. For assigned design

[ changes, NGS Engineering is responsible for identifying inputs, developing design

output and preparing the Design Change Package, including the Turnover and Close-

out checklists.

In addition, NGS Engineering is responsible for the Design Standards Program. This program develops the design bases and provides the framework for assuring that operations, including testing and maintenance, physical status and controlled documentation, are consistent with the design bases. A significant part of this responsibility is the development and maintenance of the DBDs. They are also to provide technical support to plant activities, including reviews of design changes and Safety Evaluations, and lead the development, implementation and maintenance of the Prairie Island configuration management program.

EXHIBIT A i February 11,1997 I Page 18 Part B i Rationale Ibr concluding that the design bases requirements are translated into

operating, maintenance, and testing procedures I

j Prairie Island has had processes providing reasonable assurance that design bases

! are reflected in operating, maintenance, surveillance, and testing procedures since the l plant commenced operation. Prairie Island is reasonably confident that the present j procedures address the plant's design bases. This confidence is based first on the i

plant's procedure review and control requirements, which are supported by modification

{ turnover and close-out processes. Further, system and component maintenance, j surveillances and test activities provide regular feed back on the status of the plant, i and maintain its ability to perform as designed. Lastly, the results of internal and third

! party evaluations provide additional justification for this confidence. These evaluations i have also revealed areas for improvement, for which the plant has initiated actions to j resolve, as appropriate.

l 1.0 Startup Support forPrairie Island andinitial Operation j Activities performed during construction and startup of Prairie Island provided

! reasonable assurance that the design bases requirements were correctly translated l into operating, maintenance, surveillance, and testing procedures. The original Prairie i Island Operating organization included on-site engineering personnel who were t actively involved in the development of pre-operational procedures [ pre-op), start-up

! testing, hot functionals, and operating precedures. The pre-op procedures were written i utilizing design documents such as NE provided system descriptions; vendor system j descriptions; the wiring diagrams (which had been verified by the Electrical

Constru

: tion Testing Department, with any discrepancies resolved through the NE);

associated schematics; logic and flow diagrams; and FSAR criteria.

i  !

l Preparation for initial startup resulted in a number of activities that served to improve j knowledge of the plant's design bases and confirm the integrity of its operation. These {

included the development of System Operating Procedures and the Integrated

~

l Operation Procedures. The Integrated Operations Procedures were written by the j senior members of the Operations Department. They first revised the system l

description documents to reflect the FSAR, then wrote the Integrated Operating

Procedures based on the system descriptions. These integrated Operating Procedures

were reviewed by the Engineering personnel. The System Operating Procedures were

j. written by Engineering personnel. Other procedures were similarly developed and i reviewed to support plant operation, including:

! Maintenance Procedures, which consisted mostly of component level procedures which verify the condition of the component and keep it operating

within its design parameters, which are assumed to be more conservative than

the design bases.

i l

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EXHIBIT A Febmary 11,1997 Page 19 Surveillance Procedures, which were derived primarily from the Technical Specifications, FSAR and ASME Section XI.

2.0 CommercialOperation Since their original development, procedures have been reviewed and controlled in accordance with plant instructions. Further, operational activities such as the activities of System Engineers and the conduct of surveillance, preventive maintenance, and l testing provide feedback substantiating procedural adequacy.

2.1 Technical Competence of Engineering personnel i As discussed previously, the System Engineers are a well qualified group of individuals, and one of their primary missions is assuring that system, structure and component requirements, including design bases, are understood and translated into maintenance, surveillance, testing, and operating procedures.

2.2 Surveillance Program Prairie Island has a surveillance program in place that provides reasonable assurance that SSCs governed by the Technical Specifications are capable of operating within design parameters. This includes periodic operational surveillances determined necessary for safe and reliable operation of the plant, performed within the minimum )

frequencies specified in the Technical Specifications. The Operations Committee is responsible for periodically reviewing the surveillance schedule process, and Section G of the Operations Manual, " Surveillance and Periodic Test Program." The Radiation Protection and Radiochemistry supervisors are responsible for reviewing changes to Technical Specifications, Operating and Materials Licenses for their impact on the surveillance program.

A very important surveillance test is the Integrated Si Test performed during each plant refueling outage. In this test, a manual Si accident signal is given, simultaneous with a simulated loss of off-site power [ LOOP]. The test confirms that the load sequencer performs correctly, that pumps and valves designated by design to receive an SI signal, actuate as required. It also confirms that the Emergency Diesel Generators start, reach speed and voltage in acordance with design parameters, sequence and load as required, supplying power to the safeguards busses.

2.3 Preventative Maintenance Program Maintenance practices were originally derived from vendor recommendations. An in-depth review was done in 1989 and some changes were made to more clearly address vendor maintenance requirements. Present plant maintenance procedures have benefited from accumulated plant knowledge and experience, and information

, available through the Operational Experience Assessment [OEA) Program. The preventive maintenance program is delineated in Section 11 of the Operations Manual, and is required to be reviewed periodically by the Operations Committee. Vendor design information was used to demonstrate compliance with the plant's equipment

'h i l

l EXHIBIT A 1 February 11,1997

, Page 20

! specifications, which were written to implement the plant's design bases. Since this program addressed vendor information in the development of maintenance procedures,  ;

it provides additional confidence that the design bases are being maintained.

2.4 Post Maintenance Testing Tests usually are conducted under conditions that represent normal operating '

parameters, such as flow, differential pressure, temperature, input signal values, and fluid type. In some cases, tests may be conducted under conditions that represent abnormal or emergency operating conditions. These tests assess whether the maintained component will perform as required to support defined system or structure functions.

3.0 PrairieIslandinitiatedProgram Evaluations.

In recent years Prairie Island has undertaken several self initiated evaluation programs, the results of which support our confidence in the design bases being translated into operation, maintenance, and surveillance procedures. Foremost among these was the DBD development effort. Other evaluations included program self assessments, which were focused on selected systems or activities, but whose conclusions had wider program implications.

3.1 Design Basis Document Development in the late 1980's, because of preparation for major modifications to the plant, the possibility of life extension for Prairie Island, increased regulatory and industry activity 4 in the Design Basis and Configuration Management area, and the nearing retirement of individuals with original design knowledge, Prairie Island developed a Configuration Management Program Plan. A major part of this plan was the development of the Design Bases Documents [DBDs). DBDs have been developed for the systems, structures and topical areas identified in Exhibit B.

The DBD verification effort with regard to operational activities, assessed whether:

. Surveillance and testing are adequate to demonstrate that the system will perform design bases required functions.

. System maintenance procedures support system operability, including postulated accident conditions.

. Operation procedures implement design bases requirements, including emergency and abnormal conditions.

. Training is adequate for proper operation and maintenance of the system.

This verification was done on each of the DBDs, with satisfactory results for the above operations activities. The verification process itself was validated by the performance of a SSFl on the Safety injection system, for which a DBD had previously been written and verified. A comparison of the results showed them to be very similar, thus

e 1

EXHIBIT A l Febniary 11,1997 Page 21 l validating the DBD verification effort and the plant's implementation of the design bases as they affect operations.

3.2 Setpoint Study Prairie Island's Setpoint study is an ongoing program to assess whether the Prairie  :

Island instrument setpoints are conservative with respect to the design bases. Setpoint calculations generated to date have discovered no instances of non-conservatism. ,

3.3 Surveillance Testing Reviews I In 1993, a review was performed of the plant's implementation of the surveillances identified in Section 4 of the Prairie Island Technical Specifications. All specified tests in this section were reviewed against those prescribed in plant operating procedures, with the exception of ASME Section XI inspections, which were covered in tne i IST/H10.1 program, discussed below. This review was initiated in response to a earlier  :

failure to perform a Technical Specifications required test. This review of Technical Specifications surveillance commitments identified an additional non-compliance. In context of the large number of surveillance tests specified, the results provide confidence in the thoroughness of Technical Specification surveillance testing implementation, and therefore provide reasonable assurance that surveillance  ;

procedures have not departed from the design bases.  ;

The plant's 5AWis describe the process for implementing changes to the Technical Specifications. These instructions specify reviews to determine the resulting changes '

to plant operations. In addition, the organizations affected are identified and consulted for the implementation of the change.

t The current results of Prairie Island's actions in response to NRC Generic Letter 96-01 further validates confidence that the plant implements design bases requirements. In this Generic Letter, the NRC requested utility action in response to concems that surveillances of safety related logic circuits did not reflect all Technical Specifications prescribed testing. In response to these concems, Prairie Island initiated a program to compare the plant's electrical schematic drawings and logic diagrams for the Reactor Protection System, Emergency Diesel Generator load shedding and sequencing, and actuation logic for the Engineered Safety Features Systems against the plant's  ;

surveillance test procedures. This comparison was to ensure that the safety function l portions of these systems' logic circuitry, including the parallel logic, interlocks, l bypasses and inhibit circuits, are adequately covered in surveillance procedures to fulfill the Technical Specifications surveillance requirements applicable to Prairie Island.

This review of surveillance test coverage is more rigorous than the 1993 review discussed above. Presently, approximately 95% of Unit 2's electrical circuits and 80%

of Unit 2's I&C circuits have been reviewed, with only one surveillance deficiency uncovered. This deficiency concemed the low pressure auto-start function of the Component Cooling pumps (reported under LER 96-18), which had been verified in the I

8 i

, i EXHIBIT A f February 11,1997 4

Page 22 past by other means. This deficiency was the only one identified out of the several hundred circuits / functions reviewed. l These results validate our confidence in the plant's configuration and effectiveness of the configuration management process, with respect to maintaining design bases requirements. To quote from a paragraph in 10CFR50.36 referenced in GL 96o1:

Technical Specification " surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits and that the limiting conditions of operation will be met." In addition, over the past years there have been a number of Technical Specification changes, many due to plant modifications.

The results to date confirm the ability of the plant's processes to incorporate such changes into plant procedures and maintain configuration control. i 1

3.4 SWSOPI The Service Water System Operational Performance inspection [SWSOPl] self assessment was conducted during the summer of 1995. As discussed in detail under Part C, this assessment involved over 1400 inspection hours and evaluated areas covered in NRC Inspection Procedure 40501 and Temporary Instruction 2515/1118 Revision 2. This assessment concluded that normal operating procedures support i Service Water [CL] system operation within the design basis; and that the surveillance j and testing programs established for the CL system are acceptable and will continue to i demonstrate that the system will operate as designed. The assessment concluded that l

" operator knowledge and training are excellent," and that operating procedures and  ;

surveillance and testing are " acceptable." It noted that improvements should be made  !

in post-maintenance testing.

3.5 SSFV on Safety injection A plant initiated Safety System Functional Validation [SSFV) performed by United Energy Services Corporation [UESC) in 1990 on the Safety injection System came to the conclusion that the " system will perform its intended functions and that the DBD process is being effectively implemented." Concerning plant operation, the study found that the Si system was " designed, installed, tested, operated, maintained, and i managed according to the original design basis, applicable regulations, standards codes and commitments." This study validated the DBD process, which is further described in Exhibit B.

3.6 ASME Section XI Program implementing the ASME Section XI inservice testing [lST) program required Prairie Island to determine the appropriate code classes for mechanical systems in accordance with NRC regulations and Reg Guides. This was then translated into existing or new Surveillance Procedures, which tested the mechanical components for performance in accordance with their design function. A recent comprehensive IST program review and revision have enhanced its contribution to the validation of design

J I

i EXHIBIT A i

February 11,1997 ,

Page 23 i basis compliance, and increased confidence that the IST program is performed in

! accordance with ASME Section XI.

i i The third 10-year interval IST program confirmed that the plant's ASME Code Class 1, i

2 and 3 valves and pumps can perform their safety related function, as defined in the  ;

USAR. The scope and procedural details of this program are defined in Operations j l Manual procedure H10.1. The secpe " includes pumps and valves which are required *

to perform a specific function in shutting down the reactor to the hot shutdown condition  ;

j or in mitigating the consequences of an accident." A major portion of the rewrite was i'

dedicated to providing the design basis for the components in the program. After two l years of program development, including incorporation of NRC comments, a fully developed program was put in place, and was implemented on both Prairie Island units 3

by the end of 1994. The program fulfilled the requirements of A6ME Section XI,1989 l Edition, and addressed 10CFR50.55a, Generic Letter 89-04, OM Standards, parts 1,6

and 10, and NUREG-1482.

The revised program was submitted to the NRC on September 20,1994. Additional valves requiring IST coverage were identified in LERs. The NRC responded on August 11,1995, with the third SER on the program. The response was generally favorable, with six non-generic open items. The majority of these items were resolved through  !

clarifications, and only a few necessitated changes in program details. NSP's Quality Services performed a QA audit on the implemented third ten-year IST program in July and August,1995. Though the audit report identified concerns on isolated technical issues, its conclusions were positive. It stated that the IST program was " maturing

" consistent with the expectations," was " comprehensible and clearly defined," and was

" effectively implemented." It also noted that the " engineering staff are knowledgeable of the program, regulatory and ASME Code requirements."

4.0 QA Audits and OtherEvaluations NSP's Quality Assurance organization has regularly performed QA audits and evaluations of operations activities, both procedure compliance and performance based, as discussed under Part C. These have included examination of operation, maintenance, engineering, and surveillance activities. These audits and evaluations i have generally concluded that the processes were effectively implemented, with resulting corrective actions tracked to resolution.

5.0 NRCEvaluations 5.1 EDSFI An EDSFI was conducted at Prairie Island in the spring of 1993, a detailed summary of which is provided under Part C. The NRC inspection team concluded that Prairie Island's electrical distribution system was designed, operated, and maintained in an effective manner. Specifically the following strengths were identified; competent, well qualified people with good plant knowledge to support plant operation, good '

l j EXHIBIT A i February 11,1997 i

Page 24 i configuration control programs, and excellent electrical distribution system surveillances.

1 i

5.2 SALP Assessments l Over the last 4 SALP periods, Prairie Island has received a 1 rating in the Maintenance i

Area. This area includes Surveillance and Testing. Additionally the DBD development .

' program was twice identified as a contributor for the plant obtaining a SALP 1 rating I

["exce! lent") in self assessment. Since 1990, with few exceptions, the Technical l Specifications surveillance program has been determined to be " good" or " excellent." l The Engineering discipline has consistently been classified as " good," with the System l Engineer function being frequently identified as " excellent." The SALP reports have J rated Engineering as " good" and cited a variety of incidents reflecting a less than excellent rating. SALP 11-12 did identify documentation supporting Safety Evaluations as an area worthy of improvement, and one receiving Prairie Island management attention, as discussed elsewhere.

5.3 NRCInspections Past NRC inspections have generally confirmed NSP's confidence in our operation, maintenance and surveillance program, and its compliance with design bases. NRC inspections have also noted some discrepanc;es in their past inspections, but these were generally isolated incidents that were corrected in a timely manner. Recent inspections have identified a need to reexamine some processes, particularly in the .

area of Safety Evaluation activities. Recent inspections have noted problems in the l performance of surveillances. A commitment has been docketed under a separate submittal to address these issues.

6.0 Operational Challenges to which SafeguarOFunctions Successfully Responded l

1 The ability of Prairie Island's systems, procedures and operators to successfully  ;

respond to the following safeguard challenges provides another validation of our '

confidence in the plant's design and configuration control processes, and compliance to the design bases. l 6.1 SGTR - 1979 The most notable operational challenge to Prairie Island's Safeguard's systems was the Steam Generator tube rupture, experienced on October 2,1979. The rupture was l caused by a spring on the bottom of the secondary side of the Steam Generator rubbing against the U-tubes. The RCS leak through the ruptured tube resulted in the successful actuation of the Unit 1 Safeguard systems, including Safety injection. To quote from the Licensing Event Update of February 19,1980; engineered safety systems functioned as designed and the plant operating staff accomplished safe reactor shutdown, steam generator isolation, and RCS cooldown in an expeditious i

EXHIBIT A February 11,1997 Page 25 manner following existing operating procedures." In addition, the leak and radioactive release rates "were bounded by the FSAR analysis."

6.2 Loss of Offsite Power The other safeguard challenges involve the partial or complete loss of off-site power to plant systems, resulting in the starting of the emergency diesel generators [EDGs].

The most severe challenge of this type occurred on June 29,1996, when, with both units operating at 100% power, off-site power was completely lost to Unit 2, and partially to Unit 1, due to a severe thunderstorm. As designed, the reactor and reactor coolant pumps tripped, and all four EDGs started and energized the Safeguards busses. The EDGs continued to supply power to these buses for several hours until adequate off-site power was restored. As stated in LER 96-12, "the event was mitigated by the proper functioning of the safety related equipment and the proper management of the event by the operators."

On July 15,1980, with Unit 2 at 100% power, and Unit 1 in cold shutdown, a severe electrical storm resulted in only one off-site power source being available to the Safeguards buses. As required, the reactor and reactor coolant pump tripped, and the EDGs started and energized the Safeguards busses.

On December 21 and 26,1989, due to problems with the control rod drive MG sets, the Unit 2 reactor tripped, and subsequent problems with a supply breaker for the non-safeguard buses resulted in the loss of power to them. The EDGs started, but did not energize the Safeguards busses, since they remained powered by alternate sources.

Safety injection was not actuated in these events. Though challenged, plant operators using appropriate procedures controlled plant equipment as designed. j i

I 4

i

1 EXHIBIT A j February 11,1997 Page 26 l Part C i Rationale for concluding that system, structure, and component configuration and performance are consistent with design bases

1.0 Summary Numerous construction, licensing, and startup activities provided reasonable assurance l that the plant configuration was in compliance with the design bases at initial operation.

Since initial operation, changes to the design of systems, structures and components

have been controlled through administrative controls that are part of the formal QA l program, as addressed in Part A, to maintain this reasonable assurance. Integrated in

these procedures are steps addressing compliance with design bases requirements.

j' To maintain this design change integrity, document and installation control procedures ,

include requirements for reasonably assuring that only design documents which have i been processed through the design change process are utilized for making physical

changes to plant structures, systems and components. Implementation of these procedures is confirmed by steps in the installation process and by tests, with independent oversight by the QA audit process as appropriate. These program controls were introduced in early plant ACDs, and have been built upon into the present plant SAWis.

Confidence in the Configuration Management Program has been increased through the completion and validation of the Design Basis Documents (DBD]. Other configuration management processes, such as drawing control and component data files, have been noted as contributors to the understanding and control of the plant's design bases.

Walkdowns of systems have provided an opportunity to verify the as-installed configuration of the equipment. Vertical slice assessments allow an in-depth review of design bases and operating information. Audits of processes assess the adequacy of the processes and generally support the conclusion that adequate administative controls are in place to reasonably assure that plant configuration is maintained within the design bases. Finally, NRC inspection and assessment results add to this confidence level by providing the regulator's insights to the quality of Prairie Island's 1 operation. l 2.0 Configuration Management / Design Basis Document The Configuration Management Program Plan created Design Basis Documents (DBD) for the plant systems, structures, and topical areas as listed in Exhibit B. The DBD's l are intended to capture design bases requirements and the technical approach for fulfilling them. A percentage of plant components were physically inspected to verify compliance to the design, paralleling the objectives of the NRC's Safety System Functional inspection [SSFl] Program. Discrepancies identified were entered into a l Follow-on item [FOl] process to provide appropriate review for operability and identification of corrective action. Enhanced administrative control practices were introduced so that future modifications and operational changes would be reflected in

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i l EXHIBIT A Febmary 11,1997 Page 27 l plant configuration documents, and that these changes would be reviewed against the i' now consolidated and compiled C BD's for conformance to design bases.

By the end of 1994, Prairie island had issued those DBDs that it had originally planned j as described in a November 1989 letter to the NRC. Since that time one additional DBD, for the recently installed inc:ependent Spent Fuel Storage installation (ISFSI), has j been identified as being appropriate, which is scheduled to be started in 1997. The development of these DBDs, and related activities, has resulted in the issuance of 909 3 Follow-on items (FOI), of which 93% have been closed. Approximately 80 of these i FOl's identified actions involving the SAR, six of which were determined to be l reportable and merited LERs.

3.0 Self-Assessmentinitiatives i

l 3.1 SWSOPl l The general effectiveness of Prairie Island's configuration management process, and

the knowledge of its plant system engineers, were identified in the results of the l Service Water System Operational Performance inspection [SWSOPI) self-i assessment. This self-assessment was initiated in response to NRC Generic Letter 89-13, and was conducted in accordance with NRC Inspection Procedure 40501 and

Temporary instruction 2515/1118. Preparations for the self-assessment were made by

, plant staff members. An independent self-assessment team was established, j comprising ten consultants from Yankee Engineering Services and Ogden

Environmental and Energy Services to conduct the actual assessment. Follow-up l activities are being conducted by the plant staff. To support the assessment, Prairie

Island established a five person site response team, organized around the five topical

areas identified by the NRC in their procedure

system design, operations,

. maintenance, surveillance and testing, and QA and corrective action.

i' The Yankee team expended 1,400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> in its review,1,100 at the Prairie Island plant, and issued their report on November 1,1995. They concluded that there were "no outstanding operability concems," and that the " cooling water system meets its thermal / hydraulic performance requirements." They also concluded that the " Prairie Island design change process adequately maintains the design basis and drawings

, reflect the as-constructed configuration," and stated "the process of establishing cross-functional teams ... for the entire duration of a design change from conceptional design

to operability was a significant strength." It also concluded that " operator knowledge I and training are excellent," and that operating procedures and surveillance and testing are " acceptable." Identified NSP strengths included:

. " Engineering Department Technical Knowledge" l . " Teams Established for ' Cradle to Grave' Design Changes" e " Safety Audit Review Committee" l

. " Operator Knowledge and Plant Experience" i

i l

J

EXHIBIT A February 11,1997 l Page 28

. " Communications Between Engineering, Operations, and Maintenance" l

Areas meriting improvement:

. " Flow Model Assumptions and Bases" e " Post Maintenance Testing" e "l&C Setpoints Bases, Calculations, and Control" It noted that "NSP's preparation was the best effort by any licensee. The self-assessment team did not raise an issue, ask a question, uncover a weakness or delve into matters that had not been previously identified by the response team."

The NRC based its reduced scope SWSOPl inspection on the results of this self-assessment. NSP's positive attitude and assessment preparation by the response team was addressed in that "the inspectors consider this preliminary assessment effort to be extremely proactive and indicative of an excellent safety consciousness." While stating similar positivo conclusions in its follow-up report, the NRC also had some reservations concerning the self- assessment team's attention to safety matters. It '

cited two violations concerning failure to write a Safety Evaluation for the conduct of a special test and an incorrect calculation methodology. NSP identified its corrective actions in its March 5,1996 response. These inspection results have been considered as input for continuous improvement in the Safety Evaluation process, including additional staff training on Safety Evaluations and participation in the NEl efforts on Safety Evaluations. The SWSOPl self-assessment supports a conclusion that Prairie  !

Island's control of the plant's' configuration and design basis is adequate, and that I NSP's management supports efforts that pursue a safer operating plant.

i 3.2 SSFV on Safety injection A plant initiated Safety System Functional Validation [SSFVJ performed by United Energy Services Corporation [UESC] in 1990 on the Safety injection [Sl] System came to the conclusion that the " system will perform its intended functions and that the DBD process is being effectively implemented." As stated in UESC's February 1991 report, the review was " performed to assess the functionality and operational readiness of the l Si system and to compare the results of Prairie Island's Design Basis Document (DBD)  ;

Development and Validation Program with the NRC's guidance on Safety System Functional Inspection Program. The validation was accomplished to determine whether l the system was designed, installed, tested, operated, maintained, and managed according to the original design basis, applicable regulations, standards codes and commitments."

The validation team consisted of UESC personnel with SSFI experience, and was divided into eight functional areas. The team inspection lasted from November 8 through December 12. In addition to the general positive conclusion stated above, the team " identified programmatic and technical concerns, discrepancies and unresolved

EXHIBIT A Febmary 11,1997 l Page 29 questions on 26 Requests for Information." The FOI [ Follow-on item] process was used to track the identified concems.

3.3 Station Blackout Another study that supports confidence in Prairie Island's compliance with its design 1

! bases is the DBD developed on the Station Blackout topic. The DBD was issued as l

, newly installed Emergency Diesel Generators [EDG) D5 and D6 were preparing to be integrated in the plant's systems. This DBD contained a compliance matrix of the licensing basis for this event, primarily NRC Regulatory Guide 1.155 and NUMARC 87-t 00 [the "SBO Rule"). It identified approximately eighty licensing requirements, of which all except two were shown to be satisfied. Revision 1 of the DBD documented closure j of the two issues; one because the requirement did not apply to Prairie Island, and the  !

second through implementation of a plant work instruction. Due to the complexity of this issue, Prairie Island recognized the need to establish separate license tracking  !

i mechanisms for the SBO/ESU Project. This included the issuance of the Station  !

! Blackout Rule Compliance Report in May 1989, and the development of a licensing j tracking matrix during the SBO/ESU Project.

[ 3.4 Safety Evaluations l In 1996, a self assessment was initiated on site processing of 10CFR50.59 safety

evaluations. This assessment was defined in a project definition issued on April 11,  !

1996, and consisted of three parts. The first was a review of 165 past modification and  ;

non-modification related safety evaluations, to assess the adequacy of these l l evaluations and the overall safety evaluation process, as outlined in 5AWI 3.3.2. This '

, part of the assessment was perfrmM by a team from Yankee Engineering Services i and Ogden Environmental and Eaergy ervices. The second part was an assessment  ;

i of design changes, tests and experinds that were identified as not requiring a safety l 1 evaluation. The third was an assessment using the recently developed 10CFR50.59 ,

I screening process, for the orocessing of Drawing Revision Requests (DRRs], Bypass i Requests and Authorizations, Corrective Work Orders, Non-conformance Reports

[NCRs], Follow-on items [FOls] and Technical Specifications Interpretations (TSl]. The 4

4 latter two parts were performed under the direction of an NSP engineering consultant.

Yankee Engineering issued generally positive findings on June 18,1996. They

performed their evaluation during the weeks of April 22 and May 13, utilizing 600 inspector hours, and folinwing the guidelines in NRC Inspection Manual Part 9900.

10CFR50.59, NSAC-125, and NRC Inspection Procedure 3700. Yankee stated that "the overall technical adequacy and completeness of the safety evaluations reviewed was assessed as acceptable." It also concluded that the " safety evaluation process defined by SAWI 3.3.2... meets regulatory expectations," and that "from a cumulative

perspective, the team found no evidence of safety margin erosion." The team did find

, areas of concem which merited further engineering or operability evaluations, involving

Containment Heat Removal, Containment Fan Cooler Units, Chilled Water System and

$ AFW Pump Trip. Prairie Island has initiated follow-up actions on these issues, which are being tracked as Engineering issues [Els]. The report closed by acknowledging

i I EXHIBIT A I I February 11,1997

!- Page 30 l

! Prairie Island's in process efforts to improve this area, and establish " ownership" for the 10CFR50.59 process, as discussed in the Prairie Island Plant Engineering Self-i Assessment Guide Action Plan.

I 1'

Subsequent parts of this self assessment included reviews of past modifications that were indicated as not requiring a Safety Evaluation, Drawing Revision Requests

[DRR]s, Bypass Requests and Authorizations, Corrective Work Orders, NCRs, FOls and Technical Specification Interpretations [TSis), to determine if they had warranted a Safety Evaluation. These reviews determined that under contemporary screening  ;

criteria, a more rigorous evaluation would be required in some cases. These results are to be evaluated to determine the appropriate follow-up actions, if any, particularly in l context of the planned upgrades to the Safety Evaluation procedures, SAWI 3.3.2, j discussed under Part E.

1 3.5 Plant Engineering Self Assessment i in 1995, the Prairie Island Plant Engineering Department initiated a introspective self assessment to explore ways to better fulfill its responsibilities. This self-assessment  ;

included the guidelines of INPO documents85-031, 90-009 and 92-002, and selected I guidance from NRC Inspection Manual, IP 37550 and 37551. The self assessment outline identified 23 evaluation topics and corresponding program chapters, including:

Plant Modification Control, Temporary Modification Control, Setpoint Control, Design

. Verification and Safety Review, Document Control for Configuration Management, '!

Modification Design Process, Design Change implementation, and Configuration Management.

This assessment resulted in an action plan, which identified a number of opportunities to improve the Plant Engineering department's activities that support safe and reliable plant operation. These opportunities included a variety of administrative, managerial ,

and process improvements. The status of these opportunities are being tracked for  !

completion.

4.0 Walkdowns Plant walkdowns provide another opportunity to compare the plant's physical configuration to the design documents to provide reasonable assurance that the plant's systems, structures, and components conform to design bases.

4.1 DBD Walkdown Results The DBD walkdowns and associated verification activities contribute to the reasonable assurance that configuration of plant systems, major structures, and significant topical areas covered in the DBD program are consistent with the design bases as represented in the USAR. As discussed under Exhibit B, these walkdowns checked whether component labeling and accessibility, MOV, check valve and orifice plate orientation and functional layout, were consistent with the plant drawings and system or

EXHIBIT A i Febmary 11,1997

Page 31 structure design bases requirements. Significant inconsistencies identified to date j have been resolved.

i 4.2 SWSOP1 I

As Yankee Engineering discussed in their assessment plan, the SWSOPl self-

] assessment included system walkdown for their " interconnection and interactions, as-built configuration, component layout, access for operations, inservice inspection, j maintenance, physical separation of components..." A more detailed discussion of this i effort is included in Part C Section 3.1; regarding the walkdowns, the team concluded the system was in good physical condition, and that "the drawings reflect the as-constructed configuration."

4 4.3 IE Bulletin 79-14 j A piping as-built confirmation program was initiated in 1979 at Prairie Island, to confirm

! that the configuration of safety related piping met the design requirements specified in the FSAR, as represented in the stress analysis. This program was requested by the NRC through IE Bulletin 79-14, due to identified discrepancies at several plants between the configuration of the installed piping systems and that represented in the stress analysis. For Prairie Island, actions in response to this bulletin were performed in conjunction with those for resolving IE Bulletin 79-02, concerning support / restraint anchor bolts.

The piping walkdowns were performed by Teledyne Engineering Services in accordance with their procedure P3697. This walkdown was very rigorous, including direct measurement of pipe spans, layout geometry, and valve and support / restraint locations and orientation. Also the supports / restraints were inspected for design, function and clearances. The piping configuration used in the original piping stress analysis was compared to the walkdown results. This comparison was performed by Fluor Engineers, the original plant A/E. Due to the walkdowns and as-built analysis performed during original construction, the discrepancies found at Prairie Island were relatively few. Discrepancies found during this effort were reported to the NRC and corrected. Where configuration inconsistencies were found with the original analysis, the system was re-analyzed for potentially unacceptable stress levels. The walkdowns  ;

and initial reconciliation of identified discrepancies were essentially complete by the j end of 1982. l During the mid to late 1980s, NSP revisited the IE Bulletin 79-14 efforts in preparation l for NRC inspections prior to official NRC closure of IE Bulletin 79-14 activities at Prairie Island. This involved a review of all packages for potential weaknesses in the original program. The areas revisited included documentation of valve weights and centers of gravity; check for adequate clearances where piping penetrates walls and floors (seismic rattle space); review of packages to make sure correct tolerances were used ,

for evaluating discrepancies, etc. Discrepancies were identified during these reviews. l The most notable was inaccurate weight and center of gravity documentation of Copes Vulcan control valves which were submitted through Westinghouse during original

l l

l I

EXHIBIT A February 11,1997 1 i Page 32  :

! construction of the plant. Other non-generic discrepancies were discovered and l resolved during this period. The NRC accepted this work and provided official closure i of IE Bulletin 79-14 activities.

4 l l 4.4 The Electrical and l&C Drawing Upgrade Project

, The Electrical and l&C Drawing Upgrade Project (EDUP) was initiated in 1991 in

! response to problems at other facilities and FOls generated during the DBD effort. A review of all controlled drawings for safety related electrical and l&C systems and j equipment was completed in 1995. Final drawing corrections were completed by mid i 1996. The effort involved approximately 10 person-years of NSP and contract ,

engineering resources. Approximately 4,800 drawings, involving 5,900 safety related j

components and 360 modification packages were reviewed.

! The review consisted of comparing the various drawings and plant component data file information related to each component to identify inconsistencies. Equipment j l walkdowns ware performed as needed, to resolve discrepancies. The review results l . showed that approximately 20% of the drawings contained some form of discrepancy.

! The majority of the discrepancies were typographical / informational in nature; e.g.,

component / device labeling, system codes and drawing references. The remainder

were technical in nature; e.g., wiring, contacts or incomplete incorporation of j

! modifications. Drawing corrections were made in accordance with the SAWis. '

1

! A significant portion of the technical discrepancies involved a few types of drawings. i

An example is the MCC front views which showed as-shipped layouts and labeling i

, details. This information is redundant to controlled component data files. Another i j example is low voltage panel circuit diagrams, which were essentially redundant to  ;

j wiring diagrams. Because of the duplicity of the drawings, one of the two was

! frequently overlooked when updating for modification documentation. To resolve this i situation, selected categories of drawings were deleted to eliminate redundant and

potentially erroneous sources of information. This eliminated a major source of l incomplete modification incorporation.

i

! To reduce the possibility of similar drawing discrepancies appearing in the future an j engineering guideline is in preparation. The purpose of this guideline is to disseminate

appropriate drawing conventions to the site staff to encourage drawing consistency and j- establish the set of drawings to be considered in future modifications.

i l 5.0 Audits

] NSP has maintained Quality Assurance assessment programs since initial plant

operation in 1973. These assessment activities have evaluated the adequacy of

procedures for quality related processes, and the compliance and effectiveness of their j implementation. This has included coverage of activities related to the control of plant

modifications, and their integration into plant maintenance, surveillance and operating l procedures. NSP's primary assessment activity has been the performance of QA j audits, as described in its Operational Quality Assurance Plan (OQAP). Auditing of

EXHIBIT A

~

February 11,1997 l i

Page 33 processes related to the above activities commenced at Prairie Island in August of 1973. This audit program has been supplemented at various times with other  ;

assessments, such as surveillances, in order to enhance its effectiveness. i l

Assessment activities have evaluated the adequacy of program procedures and their implementation. Inadequacies in procedures, er deviations from requirements, are documented, and corrective actions tracked 'o completion. Plant management actively participates in the definition and implementation of corrective actions.

Audits and other assessment activities are proactive with respect to addressing potential problems and industry concerns, and often go beyond regulatory requirements and commitments. Assessments utilize various approaches in order to provide thoroughness in coverage and context. This has included assessments on a process basis, assessment on a plant system basis, assessment on a modification project basis, and assessments on a issue basis. This approach has included a mixture of program and performance based audits, with at times, the inclusion of technical experts on the audit team.

These assessment activities provide additional rationale for concluding there is l reasonable assurance that: '

. System, structure, and component configuration and performance are consistent with the design bases, with deviations documented, controlled, evaluated, and corrected.

I

. Design bases requirements are translated into operating, maintenance and surveillance procedures.

6.0 ExtemalAssessments Assessments by external organizations are an important indicator of the performance of the plant staff and processes. These provide a comparison against the industry as a whole and affirm that improvement efforts are appropriately directed.

6.1 EDSFI The NRC performed a rigorous electrical distribution system functional inspection

[EDSFl] in April and May,1993, with a team that consisted of five inspectors and three technical consultants. This inspection concluded that " Prairie Island's electrical distribution system (EDS) was designed, operated and maintained in an effective manner." The inspection team identified the following strengths in system design and Prairie Island administration:

. " Robust EDS design."

e " Good plant material condition."

i l EXHIBIT A i

Febmary 11,1997 Page 34

. " Good margin in diesel generator (EDG) calculations and conservative 1

assumptions in calculations in general."

. " Competent well qualified people with good plant knowledge in sufficient ;

j numbers to support plant operation." l i e " Good configuration control program."

l

• " Excellent site EDS audits."

l e " Excellent EDS surveillances."

e " Good self assessment relative to the EDS." l l Their report also identified weaknesses, which were classified as " minor." The identified weaknesses were deficiencies in the battery design and analysis, not using i industry standards for short circuit, voltage drop and battery sizing calculations, and i

" deficiencies and inconsistencies" in the transient analysis for Emergency Diesel

. Generators D1 and D2. These items were not considered significant enough to justify a

formal response.

I The scope of the EDSFI review included the USAR, Technical Specifications and i

system design bases documents. It found the EDS design and analysis to have utilized

" proper inputs." The scope also included Prairie Island's programs for controlling modifications, alterations, temporary modifications, bypasses, engineering interfaces, drawing control,10CFR50.59 Safety Evaluations and discrepancy management. "The team concluded that Prairie Island had a good program for controlling changes or  :

modifications to plant structures, systems and components," and that " safety (10 CFR  !

50.59) evaluations were thorough and well documented." It also concluded that "each i modification and alteration was reviewed under the configuration management program i to determine its affect on the plant design basis," and that plant activities were supported by QA audits and surveillances that were "well managed," and that had "an appropriate mix of programmatic and performance based audits."

i 6.2 NRCInspection of Modification Activities l NRC inspections have provided an unique and generally favorable perspective on the )

adequacy of Prairie Island's modification activities and the actions of its managment.-

Plant modification activities, including supporting configuration management activities, have been a subject of many NRC inspections. Their conclusions have been generally positive. The NRC has recognized NSP's efforts to improve its modification control processes and to resolve weaknesses. In September and October 1990 inspections

[282/90015), the NRC noted the adequacy of the program for controlling temporary modifications and bypasses, and stated that the modification program was generally well implemented. it also stated that the QA audits of these activities were effective and in-depth. A report resulting from a January 1993 NRC inspection that focused on '

" design changes, modifications and temporary alterations," stated that "the modifications reviewed were in accordance with programmatic and regulatory requirements. Further, technical issues were appropriately addressed." This inspection resulted in no violations or deviations, and one open item [282/92026).  !

I EXHIBIT A February 11,1997 Page 35 l

The NRC's inspections have acknowledged what NSP har believed to be its typically i

aggressive stance when it comes to correcting variances from design bases commitments that have been uncovered. Regarding a deviation from a HVAC operational commitment made in the SBO Design Report [306/93015), Prairie Island d

developed an extensive matrix for all Technical Specification related essential support equipment. Similarly, when Prairie Island found that the cable separation criteria had i

been violated for a pair, the NRC commented on NSP's pro-active project which i

reviewed all main control panel cables for this situation [282-93020]. Likewise NSP's actions in response to a self-identified breach in a steam exclusion boundary were

found by the NRC to be "very aggressive", when NSP quickly established a " steam i exclusion task force," to search for other possible exclusion openings [282-93014]. In i addition, the NRC has acknowledged NSP management's continuing efforts to identify j areas in its program that may be improved, and the thoroughness of its actions. In January 1993, as result of interviews with plant personnel, the NRC noted coordination ,

l and communication " difficulties," of which it stated, the " licensee had already l l recognized this issue and was in the process of implementing specific organizational j changes," and had "recently revised the engineering manual for the system and project

engineers," which "have the potential to address" these difficulties [282/93026].

3 6.3 NRC inspection of Safety Evaluations

Though the NRC inspections and audits have generally resultad in positive

conclusions, there have been exceptions. For these exceptions, NSP believes it has

been aggressive and reasonably comprehensive in actions taken to resolve NRC j identified weaknesses in Prairie Island's program. An example of this is the NRC's i 199410CFR50.59 audit. In a report dated November 14, the NRC stated that the safety evaluations were " detailed and contained a great deal of information," and that the reviewers apparer'tly " coordinated well with other engineering groups to obtain additional information, details of analysis, independent review and expertise." The l NRC also had some negative conclusions. It stated that the evaluations lacked '

discussions on the SAR and other commitment documents, and on "the overall impact" l of design changes and "how the system function may be affected in the aggregate." It i also stated that the evaluations were "not consistently formatted." The recent NRC 1 report that transmitted the Level lli Violation resulting from a 1995 inspection recognized that " Prairie Island took brcad corrective actions to improve the 50.59 l process," and that from a technical point of view, the cited activity was adequate

[282/96015].

6.4 NRC Inspections of DBD Activities '

Several NRC inspections of the development and verification of the DBDs resulted in positive comments. The NRC report of an inspection in August and September in 1991

[282/91016], dedicated to the DBD effort, recognized Prairie Island's " determination to ensure that the design bases of the plant were identified and [then) maintained as plant modification, maintenance and operations changes occur." The NRC inspectors

T EXHIBIT A February 11,1997 l Page 36  !

concluded that the DBD project was " effective in identifying plant deficiencies." They also noted that the verification procedure "mcy not be as effective as envisioned" and l l may have compromised the quality of the verification walkdowns. NSP acknowledged l l the concern, and committed to revising the procedure and re-evaluating past l 3 verifications, actions which the NRC found to be " acceptable." Adequate instructions -

I were developed, which evolved later into several Section Work Instructions (SWis).

Further reinforcing the credibility of this program, the NRC in their July 1991 SALP ,

report commented favorably on Prairie Island's " design reconstitution" program, l referencing the verification "walkdowns and design bases document l

l [DBD] development."  !

! Later inspections further confirmed the quality of Prairie Island's DBD program. The NRC report for September through November 1992 inspection [282/92021] stated that engineering support was " excellent, as demonstrated by ... [the) design basis reconstruction activities that discovered missed surveillances..." The NRC report for the September through November 1993 inspection [282/93019), noted that a Prairie Island i

" strength was... efforts to reconstruct design basis information...." In parallel, a July 19, j 1994 inspection report [282/94008] stated that an example of Prairie Islands " good" l engineering practices was that the " plant design basis was properly utilized in performing engineering activities." The NRC's February 1996 report [282/95014) noted the completion of the reconstruction effort and issuance of the original DBD's, labeling the effort as a " good initiative."

7.0 SALP SALP reports since the inception of the DBD program in 1989 have been generally very positive with respect to management, engineering, QA and surveillance activities. The Engineering functional area has typically have been described as " good," with special mention to given to the system knowledge and the support offered by the System Engineers. Paralleling this, are the positive statements conceming NSP management's i demonstrated commitment to quality, licensing compliance and plant safety. Credit was given to NSP's management, in the July 1991 report, for establishment of the " design reconstitution" [DBD) program, mentioning the system walkdawns, design bases document development and SBO Project. With a few exceptions, the Technical Specifications surveillance program has been determined to be " good" or " excellent."

High marks have been given to QA audit and surveillance activities. SALP 11-12 identified Safety Evaluations as an area needing improvement.

t EXHIBIT A

February 11,1997 Page 37 1 Part D Processes foridentifying problems and implementation of corrective action, including actions to determine the extent of problems, action to prevent recurrence and reporting ,

i toNRC 4

l

\

1.0 Summary l

3

! Prairie Island has various problem assessment and corrective action tracking i processes, some developed specifically to meet the requirements of 10 CFR Part 50,

, Appendix B, Criteria XVI, " Corrective Action" and the Operational Quality Assurance

Plan. Operating events are addressed in accordance with the Reporting process. The

Nonconformance (NCR) process addresses items involving a " deficiency in I characteristic, documentation, or procedure which renders the quality of an item l unacceptable or indeterminate." Actions to prevent recurrence resulting from the

assessment of a nonconformance are documented and tracked with a Corrective Action i Report (CAR) through completion. The Employee Observation Report (EOR) process is used to document and correct less significant concerns. The Follow-on item (FOI)

, Process was established in accordance with the Project Plan for the Configuration

{ Management Program and used to track DBD-identified discrepancies to closure.

4

Data from the various problem identification and resolution processes are collected in a  ;

i common data base. The Self Assessment and Improvement Team (SAIT) analyzes the l

! data for adverse trends, then submits reports to plant management on a periodic basis  !

j for review and resolution. The Self Assessment and improvement Team has i

membership from most of the plant groups, including those with indirect reporting to the l Plant Manager. Safety Assessment is responsible for the functioning of the SAIT. j 2.0 Review of OperationalEvents 2.1 Operating Experience Assessment (OEA)

The OEA process is the mechanism for feedback to plant staff and corporate staff of ,

industry experience that is pertinent to plant operation, as necessary. Included in the scope of this process are NRC Bulletins and Information Notices, NRC Generic Lettes, vendor supplied experience reports,10CFR21 Reports, Prairie Island Reportable Events (RE's), INPO SER's, SO's, and SOER's, and Monticello SOE's. These events and reports are screened for applicability to Prairie Island and if appropriate are assigned to plant or corporate staff for assessment. This process provides for screening and assessing applicable operating experience documents and specifying immediate and/or corrective action as needed. Corrective actions are required to be tracked to completion.

2.2 Engineering issues The Engineering Issues (EI) database is a file to capture position papers and opinions on industry emergent events. These files are periodically reviewed by engineering to

i i

EXHIBIT A 4

February 11,1997

, Page 38 j determine applicability and priority. As appropriate, they are fed into other processes,

such as NCR, OEA, and Safety Evaluation.

l 3.0 Self Assessment Problem identification Sources i

j 3.1 Employee Observation Reports (EOR)

?

Any person noting a problem or concem can initiate an Employee Observation Report j (EOR). The Employee Observation Reporting (EOR) Process encourages employee j involvement in identifying and correcting concems observed during routine work, l rounds or other activities such as surveillances. The EOR process is a universal j problem identification and self assessment tool used to enhance work place safety, i human performance, appearance, housekeeping, equipment condition, radiation protection, environmental protection and plant reliability. This process is designed for i reporting, tracking and resolution of low threshold concems. The process includes l determining if an EOR involves a significant concem and, if so, transferring the concem l to the appropriate higher level process, such as the Nonconformance process. EORs j are reviewed regularly by the plant's Operations Committee.

3.2 Follow-on items (FOI) i The Follow-on item (FOI) process, which was developed for and used exclusively during the Configuration Management Project, established measures to guide and control the identification, evaluation, and disposition of items identified during the Design Basis

Document (DBD) verification efforts. The FOI process was designed to effectively and j promptly evaluate discrepancies for impacts on plant operability, Technical l- Specifications and reportability. The potential need for Justification for Continued

! Operation or Safety Evaluations and immediate corrective actions were addressed

through this process. A basis cause evaluation was accomplished as a part of this

{ process, which is more completely described in Exhibit B.

i i FOls were reviewed by a sub-committee of the plant Operations Committee. Corrective I l actions identified to resolve FOls are tracked to completion.

j

3.3 Plant inspection Program

The Plant inspection Program was established to identify and correct physical i i deficiencies. The intent of the program is to enhance work place appearance, safety and plant reliability by increasing manager and supervisory involvement in plant j activities. This instruction applies to the identification of concems in industrial safety, l human performance, housekeeping, equipment condition and radiation protection.  !

j improvements in specific areas are requested via the Work Control process or the EOR

process.  ;

i i

a l

4

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February 11,1997-

Page 39 i

4.0 Audit Activity l

! 4.1 Internal Audit Program i NSP maintains a written audit program to satisfy its NRC accepted Operational Quality Assurance Plan (OQAP). The following elements are incorporated in the current audit

program in accordance with the OQAP

< . Requirements for audits to be conducted in accordance with ANSI N 45.2.12 1 -1977, Requirements for Auditing of Quality Assurance Programs for Nuclear

- Power Plants; j

. Requirements for auditors to be certified in accordance with ANSI N45.2.23-1978, Qualification of Quality Assurance Program Audit Personnel for Nuclear Power Plants, as modified by Regulatory Guide 1.146, August,1980.

The audit program requires that audit subjects be scheduled such that all safety related functions are periodically audited. The audit program encompasses activities to control the plant's design bases and manage plant configuration.

5.0 Corrective Action Programs 5.1 Nonconformance(NCR)

The Nonconformance Process is established to implement the requirements of 10 CFR 50, Appendix B , Criterion XVI, " Corrective Action." A nonconformance is initiated when the quality of a part, component, system, or structure is unacceptable or indeterminate by a characteristic deficiency, a documentation deficiency, or a procedural deficiency.

Werk Orders, Receipt inspections, Engineering Issues are reviewed to determine whether a Nonconformance is required. Nonconformance operability or reportability concerns must be reported to the Shift Supervisor or Shift Manager. The nonconforming item must be controlled, and the condition must be documented on a Nonconformance Report or a Work Order (repeat or corrective).

Nonconformances are assessed by a responsible individual (usually the System Engineer) for significance and reportability, then evaluated dispositioned, and tracked.

Actions to prevent recurrence are initiated prior to closure of the Nonconformance through Corrective Action Reports (CARS). Supervisors are responsible for reviewing and closing Nonconformance reports and resulting corrective acuons (CARS), while keeping appropriate levels of management 'nformed of conditions adverse to quality.

Significant nonconformances are reviewed regularly by the plant's Operations Committee.

The Nonconformance process owner is a member of the Plant Engineering staff, who is responsible for administrative control of the Nonconformance process including monitoring and measurement.

em + _ _-- _m-_ _ ___ _-___ _

EXIIIBIT A February 11,1997 Page 40

5.2 Root Cause Evaluations

! Guidance is given in a plant SAWI on appropriate depth of root cause evaluations and

suggests various analysis techniques that may be applicable. Guidance is also provided on gathering information useful to the root cause analysis. Cause codes are provided based on INPO's Root Cause Analysis of Human Performance Events and l the Nuclear Plant Reliability Data System. These cause codes are used in identifying causes identified by various methods at the plant, including, but not limited to, the

Employee Observation Report and Nonconformance processes described above.

4

{ The root cause process is invoked by plant management for corrective action

, identification for significant events and self-assessment processes. Examples of root cause evaluation teams include the Error Reduction Task Force (ERTF) and the Self Assessment and improvement Team (SAIT).

l l 6.0 Processes llorReporting to the NRC 6.1 Reporting jv i The Reporting procedure establishes measures for the reporting of operating events to

the NRC. Site management is to be notified as soon as possible following discovery of

any abnormal or noteworthy event associated with safety related systems or activities j or essential inputs to safety related systems (all necessary attendant instrumentation, o

controls, normal and emergency electrical power sources, cooling or seal water,

lubrication or other auxiliary equipment that are required for the system, subsystem,

, train, component or device to perform its function (s) ) or any information having

potential significant implication for the public health and safety. This process addresses

reporting of items including 10CFR20.2201 Events,10CFR20.2202 Events,

{ 10CFR20.2203 Events,10CFR73.71 Events,10 CFR 50.73 Reportable Events

. (LER's),10CFR21, and 10 CFR 50.72 Events.

This process is supported by additional sub-processes that are based on NRC

! reporting regulations and available NRC guidance. Prairie Island also uses guidance contained in the TREDS process for determining reportability.

]

k 1

i i

l

o EXHIBIT A l Febmary 11,1997 i Page 41 Part E l Overall effectiveness of current processes and programs in concluding that the '

conRguration ofyourplant(s) is consistent with the design bases in controlling design l

bases con &guration 1

, 1.0 PrairieIslandResponse l t The processes and programs that have been established in accordance with the NSP

! Operational Quality Assurance Plan (OQAP) are deemed effective in providing

reasonable assurance that the configuration of the plant is maintained consistent with

! the plant's design bases; and should a discrepancy with the plant's design bases be

, identified, corrective action processes provide for appropriate resolution of the l inconsistency. The effectiveness of the Prairie Island processes has been and

continues to be evaluated through the NSP Quality Assurance Audit Program, Quality l Assurance assessments, Prairie Island self assessments, as well as assessments performed by groups independent of NSP. While these independent and self assessment activities have determined that the design basis control measures are generally effective, the Prairie Island staff has been responsive to opportunities for l improvement identified by these same assessment activities. The Prairie Island staff 4

has been self critical of the design configuration of the plant and the processes which

{ control design basis information. The Prairie Island staff strives to continue to improve

( site configuration control processes and the implementation of these processes based

] on feedback from industry issues as well as the audits and assessments performed.

2.0 Prairie Island Design and Configuration Management Controls j NSP has instituted design and configuration management controls as part of its Operational Quality Assurance Plan. These controls are included in the plant's 5AWis.

Integrated in these instructions are steps addressing compliance with the design bases requirements. These include explicit checklist items in the design input and design l verification processes. Only design documents that have satisfied these process controls are utilized in making changes to plant configuration documents. 1 i The plant's SAWis also include steps for processing revisions to operating procedures and setpoints that reflect system design changes, and provide periodic reviews that enhance fulfillment of these steps. Procurement related controls invoke parallel design and configuration controls on Architect / Engineers (A/Es], vendors and contractors that perform safety related work (and to a lesser extent Augmented Quality), and provide for monitoring of compliance. Compliance of site activities is monitored by the Generation Quality Services group. SAWl's also include provisions for evaluating, reporting and if appropriate, correcting, changes to and deviations from the SAR' requirements, in accordance with 10CFR50.59 and 10CFR50.73(a).

• SAR as used in this document refers to the FSAR and USAR. Explicit document titles are given (e.g. USAR]

when it is important for clarity, such as in describing the history of SAR changes at Prairic Island. This discussion also applies to the Independent Spent Fuel Storage Installation Safety Analysis Report [ISFSI SAR],

and to 10CFR72.48 reviews, where 10CFR50.59 reviews are discussed.

EXHIBIT A i

i February 11,1997 Page 42

3.0 PrairieIsland Self Assessments j Prairie Island maintains a competent, experienced staff of engineering personnel stationed at the site. A questioning attitude is an important component of the culture, ,

j from Site Management to entry-level personnel. Aggressive action is taken as warranted to evaluate and improve, even though Prairie Island and NSP are highlighted I by industry groups and regulators as a world-class performer. The danger of complacency is recognized. Self-assessments are regarded as a valuable tool to j validate the effectiveness of processes and practices. '

Prairie Island has been performing self assessments of the plant's design bases control, with increased focus of resources being dedicated to this area starting in approximately 1989. Most noteworthy of these activities is the Prairie Island self l

initiated Design Basis Document (DBD) Program. The Prairie Island DBD Program j provided a review of design basis information for selected systems, structures, and i topical areas. As discussed in Exhibit B the program was broad based; 16 systems, 5

structures, and 13 topical areas. The purpose of the DBD Program was to improve the i availability and retrievability of Prairie Island's design basis information. Knowing the

! extent of the DBD effort, and the configuration controls invoked on plant modifications, j there is reasonable assurance that system, structure and component configuration with j the plant's design bases will be maintained. Additional information regarding the

! Prairie Island DBD Program is provided in Exhibit B of this submittal.

l For selected systems, structures, and topical areas, the Prairie Island DBD Program l i provided for collecting and indexing of documentation supporting the plant's design  !

l basis, review and limited reconstitution of information supporting the plant's design {

! basis, and verification of the plant configuration against the design bases. The  ;

! program included creation of Design Basis Documents, resolution of open items found i during this process (FOls), and Design Basis Document verification. During this i process, discrepancies between information sources were identified. These issues were reviewed for operability, reportability and safety impact, and entered into the FOI process for corrective actions, as needed. The process has been effective in compiling

information which supports the plant's design bases and in assessing whether design j basis information has been properly incorporated into plant procedures and i configuration.

i The DBD program identified a significant number of issues for resolution. Prairie Island

believes that the program generally confirmed, in the aggregate and for the large j number of systems and topical areas reviewed, that the plant's design bases 1 supporting safe operations has been properly incorporated and maintained in plant procedures and configuration. When the program identified inconsistencies which may have posed a safety concern, the inconsistencies were promptly evaluated and reported to the NRC when required by NRC regulations. A very limited number of these issues reached the threshold for reportability, which is felt to further support the

EXHIB A Febmary n,1997 i Page 43 i effectiveness of the Prairie Island processes in providing reasonable assurance that

> the plant's configuration and operation is consistent with the design bases.

t

! Vertical slice reviews on Service Water System, Electrical Distribution System, Safety j injection System, and Safety Evaluations have identified deficiencies but few with j safety significance or indicative of significant broad based programmatic deficiencies.

j These reviews are discussed in more detail in Part B (Section 3) and Part C (Section

3).

4.0 QuaUty Assurance Assessments l' Quality Assurance assessments have been performed on key processes that provide j for the translation of the plant's design bases into documentation and physical l i configuration. Assessment subjects have included but are not limited to the following i subjects: 1) specification control, 2) safety evaluations, 3) intemal design and review, j 4) control of design inputs, 5) temporary modifications, 6) walkdowns of installed 1

! components, 7) Design Change installation, 8) Design Change pre-operational and

! operational testing, 9) setpoint control, and 10) Updated Safety Analysis Report control.

l Also, Quality Assurance assessment activities have been conducted in an attempt to be j proactive with respect to potential problems and industry concems. The above j assessments have been conducted using various approaches in an attempt to provide thoroughness in coverage. This has included assessments on a process basis, assessments on a plant system basis, assessments on a modification project basis and

{ assessments on an issue basis. Quality Assurance assessments have been performed

] via audits, surveillances and overall program assessments by organizations extemal to NSP.

4 Quality Assurance assessments have primarily been performed via NSP's Quality Assurance audit program. Auditing of processes related to configuration management j commenced at Prairie Island in 1973. NSP has continued to comply with its e commitment to perform audits as described in its Operational Quality Assurance Plan

during its operating history. Audit subjects are scheduled such that all safety related

! functions are audited. Quality Assurance audits have been performed of the Design j Change process on a periodic basis to assess the effectiveness of this key process for i controlling design configuration. These audits have determined that the Design

, Change process is generally effective and that the processes are being effectively l implemented.

l l Quality Assurance surveillances were also used for a period of time to provide on-site

monitoring of program implementation including Design Changes and Alterations. The j Quality Assurance surveillances were less structured than audits but provided i increased flexibility in day-to-day monitoring and reporting of discrepancies.

i Surveillances have been replaced by the GQS Observation Report process due to re-

engineering of the audit process in 1995.

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EXHIBIT A i February 11,1997 ,

i Page 44 l l Quality Assurance assessments have periodically identified areas of ineffectiveness, '

nonconformances or opportunities for improvement. Prairie Island views itself as j responsive to these identified issues. For example, issues were identified related to

implementation of the Design Change process in the in the late 1980s. These

identified issues did not impact the functional operability of Prairie Island but presented 3

opportunities for improvement. Prairie Island focused resources to address the l

identified issues and has continued to focus resources on maintaining an effective ,

j program for controlling changes to the plant's design bases via the Design Change i j process.

i 5.0 NRCInspections l l Inspections performed by the NRC staff have in general provided further confirmation '

i that the plant configuration and documentation has been maintained consistent with the

{ design basis to support and assure safe operation of the plant. These inspections are j discussed in more detail in Part B (Section 5) and Part C (Section 6). While these  ;

i inspections identified opportunities for improvement, they found that the inspected j l

systems were fully capable of performing their required safety functions.

6.0 Prairie Island Current initiatives j 6.1 USAR Review Project l

l Early in 1996, the NRC expressed concems over deviations from SAR requirements at

! several nuclear plants. In response, Prairie Island has undertaken an extensive USAR  !

i review program. This review program was outlined in a project plan, first issued on l May 28,1996. It established a project review team, which currently consists of seven i

permanent members, supplemented by others as expertise is required. This team j commenced their review activities in August, and have met routinely since. One  !

purpose of this review is to demonstrate that the USAR adequately describes the plant i licensing basis. Another purpose is to provided added assurance that plant's i

} surveillance, maintenance and operating procedures implement applicable design ,

bases requirements delineated in the USAR.

4 i Activities prescribed in the USAR Review plan include a review of the original plant AEC-SER to determine if necessary statements are reflected in the USAR, and a i review of USAR conformance to commitments made in correspondencs to NRC, such

! as in responses to NRC Bulletins, Appendix R submittals, and LERs. In addition, it

prescribes a review of past Design Change and Non-Mod Safety Evaluations.

Significant USAR omissions or deviations uncovered during these reviews are to be

! processed in accordance with operability, reportability, and Safety Evaluation

! requirements prescribed in the plant's SAWis. Where determined to be appropriate, j items will be processed as Nonconformances. After completion of the above, a review i of plant maintenance, surveillance and operating procedures will be performed to i determine compliance with the reconstituted USAR requirements. Conflicts are to be J

resolved utilizing existing processes.

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, EXHIBIT A 4

Febmary 11,1997 Page 45 l 6.2 Safety Evaluation Program Enhancements l The Prairie Island Safety Evaluation process has realized several recent enhancements. These enhancements have not been realized though any one action or incident, but are part of a process evolution over the past several years. Several items have fed into this process including the violations identified in NRC inspection 95014.

} The other items included inputs from the Safety Evaluation Self Assessmeni discussed in Part c., QA evaluations, and the July 1996 draft revision to the NEl Guidelines for

{ implementing 10CFR50.59, of which Prairie Island had participated in developing.

i Enhancements in Safety Evaluations have been in several forms. Formal training was

} held on the 10CFR50.59 process and SAWI 3.3.2 in the first half of 1996, with parallel j trial use of Safety Evaluation Screening (SES) criteria. These actions were i scknowledged in NRC inspection report 96015. This SES, is being incorporated into a l revision to 5AWI 3.3.2. This planned revision is to also require a written analysis confirming that a proposed change, test or experiment can be safely installed operated j~ or performed. This analysis is to be performed prior to a licensing based determination of whether the activity poses an unreviewed safety question. Further, Prairie Island

! management and the Operations Committee have an increased sensitivity to this issue.

! This is illustrated by Safety Evaluations being one of the " opportunities" for l improvement identified in the Plant Engineering Self Assessment. I Prairie Island management acknowledges the NRC's recent Level lli Violation and l recognizes that further appraisal of the Safety Evaluation process is appropriate.

i Prairie Island believes that that the upcoming revision to SAWI 3.3.2, recent training

and other internal actions are very significant steps toward bringing the Safety Evaluation process to within contemporary expectations.

I 7.0 Conclusion j The objective evidence provided by inspections and independent assessment activities, l Quality Assurance assessments and other self assessments supports the conclusion

. that the Prairie Island processes are effective and are effectively implemented to

! provide reasonable assurance that the configuration and operation of the plant is

' consistent with the design bases to assure safe plant operation. The assessments performed also indicate that the Prairie Island staff is self critical with regard to maintaining and operating the plant safely. The Prairie Island staff has used the site corrective action processes to self identify issues, assess the significance and

implement appropriate corrective actions.

it is also noted that regulatory requirements and quality expectations have changed

significantly over the life of the facility. Prairie Island recognizes that program and

process effectiveness is only one indicator of performance. The Prairie Island sta'f

continually seeks improvements in the controlling processes and process 1

implementation to provide further enhancements such that the facility is maintained and e

EXHIBIT A February 11,1997 Page 46 operated in the most effective manner feasible, while maintaining the primary focus on public health and safety, as well as compliance with regulatory requirements.

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4 EXHIBIT A February 11,1997 I

Page 47 i ,

l References

Prairie Island 5AWI Work Instructions- -

1 5A WI 1.5.0 Procedure Control 5AWI 1.5.1 Procedure Deviation I 5A W I1.5.2 Procedure Submittal Process 5A W I1.5.3 Periodic Procedure & Checklist Review i SA WI1.7.0 Control Of Commitments & Documents Control Of NRC Commitments

SA W I1.7.1 5A W I 2.1.0 Quality Assurance Program Boundary i SA WI 3.2.0 Work Control 5AWI 3.2.1 Work Control Package initiation SAWI 3.2.2 Work Control Package Preparation & Review 5A WI 3.2.4 Conduct Of Work 5A WI 3.2.5 Completed Work Package Reviews i 5A WI 3.3.0 Operations Committee 1 5AWI 3.3.1 Technical Specification Change Review Committee l 3

5AWI 3.3.2 Safety Evaluations 5A W I 3.4.0 QA Records Control 5A WI 3.6.0 Reporting .

SA WI 3.6.3 10CFR & Tech Spec Reporting Requirements  !

SA WI 3.6.5 Determination Of 10CFR21 Reportability l 5AWI 3.7.0 Operating Experience Assessment 5A WI 3.9.0 Bypass Control 5AWI 3.10.0 Control & Operation Of Plant Equipment SAWI 3.10.5 Plant Equipment Labeling SAWI 3.12.0 Nuclear Plant Maintenance 5AWI 3.12.4 Post-Maintenance Testing 5AWI 3.14.1 Setpoint Control SAWI 3.15.2 Employee Observation Reporting SAWI 3.15.3 Plant inspection Program (PIP) 5AWI 3.15.5 Operability Determinations SAWI 3.16.0 Plant Surveillance Program SAWI 3.17.0 Site Trending Program SAWI 3.17.1 Root Cause Evaluation & Corrective Action Guidance 5AWI 3.20.0 Testing SAWI 3.21.0 Certification Of Inspection, Examination, & Testing Personnel SA WI 4.1.0 Plant Operations & Site Engineering Manuals 5A WI 4.1.1 Maintenance Of Pi Analysis Log & Index Data Base SA WI 4.4.0 Drawing Control 5AWI 4.4.1 Controlled Drawing Files

1 i EXHIBIT A February 11,1997

Page 48 j 3

5A WI 4.4.2 Site Controlled Drawings 5A WI4.4.3 Transitional Drawing Files SA WI4.4.4 Drawing Additions, Revisions, & Deletions

! 5A WI 4.5.0 Plant Component bats Files i 5A WI 4.6.0 Specification Control

} 5A WI 4.9.0 Safety Analysis Reports

! SA WI 6.0.0 Integrated Planning Process l 5A WI 6.1.0 Design Change General

! 5A WI 6.1.1 Design inputs

5A WI 6.1.2 Design Outputs, Review & Verification 5AWI 6.1.3 - Design Change Package, Logs & Records 5A WI 6.1.4 Design Change Description / Safety Assessment 5AWI 6.1.5 Design Change implementation Plans

, 5A W I 6.1.6 Design Change Review & Approval 5AWI 6.1.7 Design Change Work Orders ,

j 5A WI 6.1.8 Engineering Change Requests i

SAWI 6.1.9 Design Change Tumover For Operation

} 5AWI 6.1.10 Design Change Close-Out 5AWI 6.1.19 Site Alteration Process SA WI 6.2.0 Substitute Part / Component Equivalency (SPCE) Evaluation SA W I 6.5.0 Temporary Modifications  !

SAWI 7.1.0 Site Procurement Process l

5A W 1 7.1.1 Procurement Planning Process j SAWI 7.1.2 Nuclear Grade Procurement j 5A W I 7.1.4 Augmented Quality Procurement i 5AWI 8.1.0 Receiving Process

SA W I 8.2.2 Control Of Nonconforming items & ltems Removed From

! Service

5A W I 8.4.0 Nonconformances

[

SAWI 14.5.0 Site Inspection Program

, NRC & NRR Evaluations:

NRC Inspection Report number 282/90015; 306/90016: routine, announced l inspection of design changes and modifications, conducted on September 17 through October 5,1990.

NRC Inspection Report number 91001: SALP 9 Report, covering the period of

!_ May 1,1989 through April 30,1991

)

i NRC Inspection Report number 91016: routine, announced inspection of

licensee Design Basis Document Verification (DBDV) program implementation, j conducted on August 26 - 30, and September 11,1991.

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! EXHIBIT A February 11,1997 Page 49

NRC Inspection Report number 92021. routine unannounced inspection of j outage activities, including maintenance, Surveillance and self assessment,

conducted on September 15 through November 9,1992.

j NRC Inspection Report number 92026: safety inspection of design changes, l modifications and temporary alterations, conducted on January 11 - 22,1993.

. NRC Inspection Report number 93007: special announced EDSFI in i

accordance with Temporary Instruction 2515/107 (25107) conducted on April 12

{ through May 14,1993.

NRC Inspection Report number 93014: routine, unannounced inspection by resident and regional inspectors of plant operational safety including onsite follow-up of events, maintenance, surveillance, engineering and technical support, conducted on July 20 through September 13,1993..

NRC Inspection Report number 93015: special safety inspection relating to D5/D6 switchgear and HVAC, conducted on July 19 through 30,1993.

NRR audit on implementation of 10CFR50.59, conducted on August 8 -12,1994.

NRC Inspection Report number 93019, routine unannounced inspection of operational activities, including maintenance, surveillance, engineering and technical support, conducted on September 14 through November 8,1993.

NRC Inspection Report number 95014: integrated routine inspection of  !

operations, maintenanco, engineering and plant support, conducted on November 21,1995 through January 2,1996.

SALP 12 Report, covering the period of July 24,1994 through February 17, i 1996 NRC Inspection Report number 96015: inspection to determine if the licensee's proposal to take credit for the non-seismic intake canal and operator actions following an earthquake constituted an unreviewed safety question, conducted on December 22,1995 through October 9,1996.

EXHIBIT A

! February 11,1997

! Page 50 i

Acronyms 5AWI - Plant Administrative Work Instruction

] NE - Architect Engineer ACD - Administrative Control Directive l AFW- Auxiliary Feed Water 1

ASME - American Society of Mechanical Engineers

CAR - Corrective Action Report l CCFS - Calibration Card / Setpoint File i DBD - Design Basis Documents

DRR - Drawing Revision Request

[ ECR - Engineering Change Request j EDG - Emergency Diesel Generator

. EDS - Electrical Distribution System j EDSFI - Electrical Distribution System Functional inspection j EDUP - Electrical and l&C Drawing Upgrade Project l El- Engineering issue

! EOR - Employee Observation Report

EQ - Environmental Qualification

ERTF - Error Reduction Task Force l ESU - Electrical System Upgrade 4

FOI - Follow-on item

FSAR - Final Safety Analysis Report i GQS - Generation Quality Services i l&C -Instrument and Control

IE -Inspection and Enforcement

! INPO - Institute of Nuclear Power Operations

IST-Inservice Testing i LER - License Event Report MCC - Motor Control Center a N1 ACD - Corporate Nuclear Administrative Control Directives

! NAD - Nuclear Analysis and Design NCR - Nonconformance

, NDE - Non-destructive Examination l NEl- Nuclear Energy Institute

NGS - Nuclear Generation Services i NRC - Nuclear Regulatory Commission i

NSAC -

i NSP - Northern States Power l NUMARC - Nuclear Utility Managements and Resources Council

OC - Operations Committee j OEA - Operational Experience Assessment

} OM OQAP - Operational Quality Assurance Plan 2

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EXHIBIT A February 11,1997

, Page 51 PDISA - Project Description / Safety Assessment l PP - Project Plan

. QA- Quality Assurance 4

QC - Quality Control RCS - Reactor Coolant System SAIT - Self Assessment and Improvement Team i SALP - Systematic Assessment of Licensee Performance

, SAC - Safety Audit Committee SAR - Safety Analysis Report SBO - Station Black Out SER - Safety Evaluation Report i 3

SGTR - Steam Generator Tube Rupture l SI- Safety injection  !

4 SPCR - Setpoint Change Request SRO - Senior Reactor Operator l SSC - Structure, System, and Component SSFI - Safety System Finctional Inspection SSFV - Safety System Functional Verification SW1- Section Work Instruction  !

i SWSOPI - Service Water System Operational Performance inspection  !

TMl-Three Mile Island i

T-Mod -Temporary Modification '

i TSI - Technical Specification Interpretation

{ USAR - Updated Safety Analysis Report J

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EXHIBIT B  !

l Febmary 11,1997 l l Page1 Exhibit B - Configuration Management Program In 1989, NSP initiated a more rigorous Configuration Management Program for the  ;

Prairie island plant, which included the development of Design Bases Documents I j [DBDs) to identify the design bases for selected plant systems, structures, and topical i areas as listed in the table at the end of this Exhibit. This program was outlined in j NSP's November 13,1989 letter to Region lil of the NRC. The DBDs are intended to j capture design commitments delineated in the SAR documents, and the technical

, approach for fulfilling them. Plant configuration in addition to maintenance and j

operating procedures were reviewed to verify ther.e DBDs, paralleling the objectives of

{ the NRC's Safety System Functional inspection (SSFl] Program. Identified j discrepancies have been entered into the Follow-on item [FOl] procers to provide j appropriate corrective action. Enhanced document and design control practices were

! introduced so that future modifications and operational changes would be reflected in

! plant configuration documents and plant procedures, and that theso changes would be l consistent with the now consolidated and compiled Design Basis Documents (DBD].

i The NRC was kept informed of the progress of this effort, and several LERs were issued identifying a reportable item. By the end of 1994, NSP issued and verified DBDs for systems, structures and topical areas as identified in the 1989 letter to the NRC.

P!snt work instructions were developed to implement this configuration management program. Those relating to modification activities are discussed in Part A of Exhibit A to this response. The present version of the primary work instructions for the DBD effort are described below-Preparing Design Bases Documents. A plant work instruction defines the methodology for preparing and reviewing DBDs for Prairie Island systems, structures and topical areas. DBDs were developed by the Design Standards group, under the direction of a Supervising Engineer. The primary participants in its development were the Responsible Engineer, the Responsible Review Engineer, and the DBD Review Team. The DBD Review Team was generally comprised of the System Engineer and other persons knowledgeable in the DBD subject area.

As discussed in this instruction, the primary content of the DBD are the technical requirements upon which the system, structure or topical position is based. The scope of these requirements parallels the definition of design inputs provided in ANSI N45.2.11. These requirements include regulatory commitments, a principle source of which is the SAR. Other commitments are derived from Safety Evaluation Reports, Branch Technical Positions, draft 1967 AEC General Design Criteria, NUREGs, and correspondence between NSP and the NRC. Often these commitments are reflected in Regule'ory Guides, and industry codes and standards, which are referenced in the DBDs. In addition to the regulatory commitments, DBDs may include specific design requirements, such as:

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EXHIBIT B

! February 11,1997 Page 2

e The system or component's function, including severity and frequency of i

accident cenditions;.

. Performance requirements or limits; t

. Design parameters, such as pressure, temperature or voltage; I e Loads, such as thermal, wind or seismic conditions; )

e Environmental conditions; such as local humidity and temperature conditions; ,

i e Interface requirements; l

• Physical and radiological protection barners; e Mechanical, electrical, structural and instrumentation & control requirements;.

. System separation, redundancy and fire protection requirements;

. Operations requirements, such as plant startup, normal operation , shutdown  ;

l and emergency operation; and i j e Maintenance requirements, i

The instruction specifies that a number of source documents are to be considered in developing the above requirements. Besides the regulatory related documents  !

j mentioned above, these include calculations, vendor manuals, plant drawings, l

l modification pec ages, engineering evaluations and analysis, specifications, plant j

{ operation procedures, procurement documents, and the plant's Technical l

! Specifications. It also specifies that where source documents are not available through I l the plant's microfilm or controlled manual, a DBD Source Document Data Sheet shall

be completed. This instruction also requires that a DBD Boundary Definition be developed, which shall identify the physical boundaries of the system or structure, and  ;

its functionalinterfaces. '

1 l The draft of DBDs and DBD Boundary Definitions were required to receive a detailed

review by the Responsible Revievi Engineer and the DBD Review Team. Upon

! resolution or notation of the comments, a DBD was issued for Trial Use in the verification process defined in SWI ENG-4, discussed below. After completion of the verification and resolution of its results, a DBD is issued for use as Revision 0, and is formally controlled.

Verifying Design Bases Documents. A plant work instruction defines the process for ensuring that the system or structure configuration, surveillance, maintenance and operation conform with the requirements outlined in the DBDs. A DBD Verification Package was developed for each DBD, and was initially issued as Revision 0, for implementation. Component importance rankings, based on the component's )

importance to safety, were obtained from the Nuclear Analysis Department, to aid in selecting a vertical slice representative of the system / structure configuration and for focusing the verification activities. The DBD Verification Package contains a columnar checklist, which was broken into the following sections, based on the major verification topics:

EXHIBIT B j February 11,1997 i Page 3 l . Design Bases Requirements Verification: covers specific design bases j requirements that are to be satisfied during accident or abnormal conditions.

l

. Design Verification: covers design bases requirements that are to be

incorporated in the design.

! e Maintenance and Testing Verification: covers maintenance and testing activities

to assure that system and components will perform their safety function, and that j maintenance activities will not compromise this function.

! . Walkdown Verification: physical verification that the physical configuration in the

{ plant are in agreement with the drign bases.

j e Procedure Verification: operatbnal procedures are consistent with design bases l requirements, including training and component surveillances, i e Operational Experience Verification: verifies that plant and industry operating l experience for a system or component is factored into its design, operation and j maintenance.

l l The system walkdown verification sought to verify (through sampling) the correct

component labeling and accessibility, and correct MOV, check valve and orifice plate

! orientation. It included a system line-up verification to confirm that plant procedures l and drawings are consistent with as-built plant conditions and design bases

requirements. This latter effort involved a comparison of a sample of the components i

in piping systems against their installation isometrics. The system DBD checklists also 1 i included subsections for a sampling of valves, pumps, instrumentation and sensors, I

, and circuit breakers / isolation devices. l l

l The instruction requires independent review of the DBD Verification Package. Tl;e

completed package was used by the DBD Verification Team, which was led by a assigned Team Leader. The team had access to plant personnel, as necessary to fulfill their responsibilities. The instruction states that the team is to be guided by the  ;

DBD Verification Package, but not restricted by it. The checklist provided columns for i recording element acceptability, references and bases for the conclusions, and comments or clarifications. After completion of a DBD verification, the DBD Verification Package was revised and issued. Items found to be inconsistent with the DBD, or that i required future follow-up actions to determine acceptability were identified in FOls.

I Identifying, Asssssing, and Resolving Configuration Management Related Follow-On items. Items that "may represent a difference or conflict between the as-built plant configuration", plant configuration documents or the design bases /DBDs, or that may represent a difference between the DBD and the source design bases document were identified in FOls. Initiators of FOls utilized form PINGP 1152, entering as much related information as possible. A plant work instruction requires the Design Standards Supervisor to perform an initial screening of FOls, and those that "may include operability impacts, or most likely impact significant plant issues" are to be classified as Priority Follow-On items. Those that are simple or repetitive are classified as Routine

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! EXHIBIT B

February 11,'1997 j- Page 4 i

Follow-On items [RFOls). The Design Standards Supervisor, or his designee, is also

l. required to log in the FOls, providing for future tracking.

After initial screening and logging, the instruction requires that the assigned Responsible Engineer perform a detailed assessment of the FOl, for determining the

appropriate actions to resolve the item and prevent future recurrence. This evaluation j is to be recorded on forms PINGP1152 or 1157, depending on its operability impact.

For those that are determined not to represent a conflict with the plant design bases are classified as not valid, and no corrective actions are required. For valid and non-

! routine FOls, the Responsible Engineer is to perform a " prompt assessment to I j determine if there are concerns." The criteria for this assessment follow.

! . Operability Assessment: determine if the FOI" impacts the operability of plant i systems, structures or components," following the guidance of SAWI 3.15.5,

" Operability Determinations". This includes determining the quality and non-j quality related functions of the subject item, and whether there is an impact on

the item's ability to perform that function.

e Technical Specifications / Plant Commitment Assessment

determine if the FOI j affects equipment covered in the plant's Technical Specifications or other plant i specific commitments.

. Reportability Assessment: assess whether the FOl is reportable to the NRC in accordance with 10CFR21 or 10CFR50.72, merits a Licensing Event Report i (LER), or is reportable under another criteria. This assessment and the resulting

)' reporting are to be performed following the applicab!a 5ACDs and SAWis.

i e Safety Evaluation / Justification for Continued Operations Assessment: determine i if the FOI " involves a change in the facility or its operation as described in the

SAR, and thus requires a Safety Evaluation. If the result is positive, and the

FOI also " involves a change in the plant's Technical Specifications or an

! unreviewed safety question," per 10CFR50.59, it also requires a JCO.

t i

j The results of this assessment are recorded on form PINGP 1157. The Responsible

! Engineer is required to bring FOls that raise operability or Technical Specifications

! concerns to the immediate attention of the Design Standards Supervisor, for follow-up i with plant management. FOls that are determined to impact operability, Technical Specifications, are repostable, require a JCO/SE, or merit immediate corrective action are to brought to the attention of plant management. The final disposition of the FOI in regards to these issues is to be recorded on the form PINGP 1157.

This instruction requires that the Responsible Engineer evaluate all valid non-routine FOls to determine the root cause of the item; the engineer records the results on form PINGP 1157, and initiates the appropriate corrective action. This evaluation is to also determine if an NCR is required, fcilowing the guidance of Tables in 5AWI 8.4.0, and initiate the same if appropriate. FOls are subsequently to be categorized based on the immediacy of their corrective actions, and issued for action. Long term corrective actions are to be tracked by the Design Standards Supervisor, or his designee.

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EXHIBIT B February 11,1997 Page 5 As of the end of 1994, the DBDs NSP had committed to complete in its November 1989 letter to the NRC were issued. Since that time one additional DBD has been identified as being appropriate. It is scheduled to begin in 1997. Attachment 1 is list of all issued DBDs for Prairie Island.

The development of the DDDs, and related activities, resulted in 909 fob, of which approximately 93% have been closed. Approximately 80 of these FOls involved conflicts with the SAR, and 6 were determined to be reportable, and merited LERs.

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EXHIBIT B February 11,1997 Page 6 DBD Number DBD Title DBD-STR-01 CONTAINMENT DBD-STR-02 AUXILIARY BlJILDING DBD-STR-03 TURBINE BUILDING DBD-STR-04 SCREENHOUSE BUILDING DBD-STR-05 D5/D6 BUILDING DBD-SYS-04 REACTOR COOLANT SYSTEM DBD-SYS-08 REACTOR PROTECTION SYSTEM DBD-SYS-12A CHEMICAL VOLUME AND CONTROL SYSTEM DBD-SYS-14 COMPONENT COOLING SYSTEM DBD-SYS-15 RESIDUAL HEAT REMOVAL SYSTEM DBD-SYS-16 SPENT FUEL POOL COOLING SYSTEM DBD-SYS-18A SAFETY INJECTION SYSTEM DBD-SYS-18D CONTAINMENT SPRAY SYSTEM DBD-SYS-20.05 4KV MISCELLANEOUS AUXILIARIES SYSTEM DBD-SYS-20.06 480V MISCELLANEOUS AUXlLIARIES SYSTEM DBD-SYS-20.09 DC AUXILIARIES SYSTEM DBD-SYS-20.10 240/120 VOLT MISCELLANEOUS AUXILIARIES DBD-SYS-28B AUXILIARY FEEDWATER SYSTEM DBD-SYS-34 STATION AND INSTRUMENT AIR SYSTEM DBD-SYS-35 COOLING WATER SYSTEM DBD-SYS-38A DIESEL GENERATORS SYSTEM DBD-TOP-01 ACCIDENT ANALYSIS DBD-TOP-03 ENV QUAL OF ELECTRICAL EQUIPMENT DBD-TOP-04 NUREG-0737 DBD-TOP-05 HAZARDS DBD-TOP-06 FIRE PROTECTION / APPENDIX R DBD-TOP-07 FUEL HANDLING DBD-TOP-08 CONTROL OF HEAVY LOADS DBD-TOP-10 MOTOR OPERATED VALVES DBD-TOP-11 PLANT PIPING AND SUPPORTS DBD-TOP-12 REG GUIDE 1.143 COMPLIANCE DBD-TOP-13 SElSMIC REQUIREMENTS DBD-TOP-15 STATION BLACKOUT DBD-TOP-16 ELECTRICAL DESIGN ISSUES