ML18153A949

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Summary of 940422 Meeting W/Util in Atlanta,Ga Re Util Engineering Support to Plants.Nrc of Opinion That Meeting Beneficial in That Meeting Provided Better Understanding of Corporate,Site & Status of Selected Engineering Activities
ML18153A949
Person / Time
Site: Surry, North Anna  
Issue date: 04/28/1994
From: Verrelli D
NRC Office of Inspection & Enforcement (IE Region II)
To: Stewart W
Virginia Power (Virginia Electric & Power Co)
References
NUDOCS 9405230023
Download: ML18153A949 (56)
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Docket Nos. 50-280, 50-281, 50-338, 50-339 License Nos. DPR-32, DPR-37, NPF-4, NPF-7 Virginia Electric and Power Company ATTN:

Mr. W. L. Stewart Senior Vice President - Nuclear 5000 Dominion Boulevard Glen Allen, VA 23060 Gentlemen:

SUBJECT:

MEETING

SUMMARY

- ENGINEERING UPDATE AND INITIATIVES

,This refers to the meeting conducted at your request at the NRC Region II Office in Atlanta, Georgia on April 22, 1994.

The meeting's purpose was to discuss Virginia Electric and Power Company's engineering support to the North Anna and Surry nuclear power plants~

It is our opinion that this meeting was beneficial in that it provided us with a better underitanding of the corporate, site, and system engineering support to and status of selected engineering activities and initiatives. Specific topics included system engineering, backlog reporting, five year license renewal, equipment data system, nuclear analysis capability, Surry core uprate, and design basis documentation.

In accordance with Section 2.790 of the NRC's "Rules of Practice," Part 2, Title 10 Code of Federal Regulati-0ns, a copy of this letter and its enclosures will be placed in the NRC Public Document Room.

Should you have any questions concerning this letter, please contact us.

Enclosures:

1. List of Attendees
2.

Engineering Update and Initiatives cc w/encls:

(See page 2) 9405230023 940428 PDR ADOCK 05000280 P

PDR Sincerely, j~ r 1 1~' }iildlL, David M. Verrelli, Chief Reactor Projects Branch 2 Dlvision of Reactor Projects

Virginia Electric and Power Company cc w/encls:

M. L. Bowling, Jr., Manager Nuclear Licensing Virginia Electric & Power Company 5000 Dominion Boulevard Glen Allen, VA 23060 G. E. Kane, Station Manager North Anna Power Station P. o~ Box 402 Mineral, VA 23117 M. R. Kansler Station Manager Surry Power Station P. 0. Box 315 Surry, VA 23883 Executive Vice President Old Dominion Electric Cooperative 4201 Dominion Boulevard Glen Allen, VA 23060 Dr. W. T. Lough Virginia Corporation Commission Division of Energy Regulation P. 0. Box 1197-Richmond, VA 23209 William C. Porter, Jr.

County Administrator Louisa County P. 0. Box 160 Louisa, VA 23093 Ray D. Peace, Chairman Surry County Board of Supervisors P. 0. Box 130 Dendron, VA 23839 Michael W. Maupin, Esq.

Hunton and Williams Riverfront Plaza, East Tower 951 E. Byrd Street Richmond, VA 23219 cc w/encls cont'd:

(See page 3) 2

/'1

Virginia Electric and Power Com*pany cc w/encls cont'd:

Attorney General Supreme Court Building 101 North 8th*street Richmond, VA 23219 Robert B. Strobe, M.D., M.P.H.

State Health Commissioner Office of the Commissioner Virginia Department of Health P.O. Box 2448 Richmond, VA 23218 bee w/encls:

G. Bel isle, RII L. Garner, RII B.. Bue kl ey, NRR L. Engle, NRR Document Control Desk NRC Resident Inspector U.S. Nuclear Regulatory Commission Route 2, Box 78-A Mineral, VA 23117 NRC Resident Inspector U.S. Nuclear Regulatory Commission Surry Nuclear Power Station 5850 Hog Island Road Surry, VA 23883 RII :DRP l-u>G--

LGarner 4/)4/94 RII:DRP Rit\\DI,_~

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ENCLOSURE 1 LIST OF ATTENDEES A. F. Gibson, Director, Division of Reactor Safety (DRS), Region II (RI!)

T. A. Peebles, Chief, Operations Branch, DRS, RII P. J~ Kellogg, Chief, Operational Programs Section, DRS, RI!

J. R. Johnson, Acting Director, Division of Reactor Projects (DRP), RI!

D. M. Verrelli, Chief, Reactor Projects Branch 2, DRP, RI!

M. W. Branch, Senior Resident Inspector - Surry; DRP, RII D. R. Taylor, Resident Inspector - North Anna, DRP, RI!

L. W. Garner, Project Eng~nee~, Reactor Projects Section 2A, DRP, RI!

H. N. Berkow, Director, Project Directorati II-2, Office of Nu~lear Reactor Regulation (NRR)

B. C. Buckley, Senior Project Manager, Project Directorate II-2, NRR L. B. Engle, Project Manager, Project Directorate II-2, NRR Virginia Electric and Pow~r Company E.W. Harrell, Vice President, Nuclear Engineering Services D. L. Benson, Manager, Nuclear Engineering, Nuclear Engineering Services E. S. Grecheck, Manager, Inservice Inspection/Nonde~tructive Examination and Engineering Progra~s, Nuclear Engineering Setvices K. L. Basehore, Supervisor, Nuclear Safety Analysis, Nuclear Analysis and Fuel, Nuclear Engineering Services M. L. Bowling, Manager, Nuclear Licensing Program, Nuclear Services

Engineering Update And Initiatives April 22, 1994

................................. VIRGl/f/lAPOWER PT11M.1.

E. W. Harrell Vice President Nuclear Engineering Services PT11M.2

Agenda Introduction E.W. Harrel I System Engineering, Backlog Reporting

  • D.L. Benson Five Year License Renewal E.S. Grecheck Equipment Data System E.S. Grecheck Nuclear Analysis Capability K.L. Basehore.

Surry Core Uprate K.L. Basehore Design Basis Documentation D.L. Benson

  • Summary E.W. Harrell PT11M.3

Major Engineering Activities And Initiatives I

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  • Station Support
  • Reload Fuel Design
  • Ten Year ISi
  • Technical Specification Development
  • ,
  • Modifications

- Station Blackout

- RTD By-Pass Removal

  • Regulatory Compliance

- IPE/IPEEE

- SQUG

  • Engineering Initiatives T.S./UFSAR Surveillance Review Five Year License Renewal PT194,4

North Anna Unit 2 Steam. Generator Replacement Status April 1994

  • Same Project Team
  • Re,placement Scheduled For Fall 1996

- Delivery On-Site, February 199 5

  • Design Engineering (DCP) - 40°/o Complete

. - Scheduled DCP Approval, February 1995

  • Long Lead Material - 1 0°/o Complete

- Materials Scheduled.On-Site, May 1995 PT194.5

Nuclear Engineering Services Organization VP Nuclelr En~rH!llrlng S.VIDIII E.W.Hlffflll I

MalllglK Manager Project Mln11ger Manager Nut1.. Anuyafs II Full 181/NDE a St*m Gentmor R.M. l!ooynn En~nearlng P119111111 RepllallTl8lll Proj9d Nut1M ~r-'ng It Baehore E.9. Gnlchldl.

(North Anna)

D.L Binion I

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I Supervl1or Supervl1or Supervisor Supervl1<<

Supednt.ndenl

&.opednllndlnt ISIINDE ISIINDE Proc:t.1W118111 PIOCUf9fflllll a Engl,-t1111 Programs

& Engl...tng P119111111 EnglnNrlng En~neerlng Englnnrlng Engln*rlng (Sooy)

(North Anna)

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(North Anna)

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(North Anna)

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Manager Managlf ChllMlchancll Nilda' s.drlcll

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System Engineering Backlog Reporting D.L. Benson.

Manager Nuclear Engineering PS1D4.8A

Engineering Backlog Reporting The Following Backlogs Are Assessed On A Monthly Basis

  • Design Change Requests
  • Design Change Implementation
  • Drawing Backlog
  • Drawing Production
  • Justifications For Continued Operation
  • Commitment Tracking Items
  • Station Deviation Reports
  • Equipment Data System Parameter Set Upgrades
  • Engineering Standards And Procedures Reviews
  • Potential Problem Reports
  • Vendor Technical Manual Change Requests
  • Design Basis Document Open Item Closeout

.

  • Engineering Projects, Types 1, 2 And 3 I

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PT11M.7

Virginia Power Initial Five Year Term License Renewal Initiative E. S. Grecheck Manager Inservice Inspection / Nondestructive Examination And Engineering Programs PT11M.8

Virginia Power 5-Year License Renewal Rationale I

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  • Intended To Complement Twenty Year Term
  • 5 Year Application Should Be Less Complex

- No Limiting Environmental Issues

~ No Known Technical Limitations

  • Cost Of Application Is Minimized And Expected Economic Benefits Are Significant
  • Should Reduce The Likelihood Of Public And Local/State Government Objections PT11M.II

Virginia Power 5 Year License

. Renewal Benefits

  • Reduces Near-Term Decommissioning Trust Fund Collections
  • Reduces Need For Rate Increases
  • Benefits Industry, May Preclude Premature Shutdowns
  • Provides More Incentive To Invest In Plant And Technological Upgrades To Maintain Higher Availability PT11M.10

Virginia Power License Renewal Study Phase I

  • Economic Benefit Study
  • Review Surry PLEX Reports
  • Determine Licensing Strategy
  • Perform Preliminary Environmental Evaluation
  • Coordinate With NEI And Owners Groups
  • Develop Final Scope of Work, Cost Estimates, And Schedules Phase II

Licensing Strategy

  • Use Industry Developed Process For Performing The IPA
  • Use The Revised 1 OCFR 51 And 1 OCFR 54 Rules When Issued
  • Exempt Any 1 OCFR 54 Requirement That Plant Must Be Greater Than 20 Years Old.

To Submit A Renewal Application

-.1 PT194.12

Integrated Plant Assessment Process

  • Objectives

- Simplify Documentation Requirements And Reduce The Need For Additional Analysis

- Provide The Information Needed By The NRC To

-Issue Renewed Licenses

- Eliminate/Minimize Process Complexity

- Minimize Costs PT1114.13

Integrated Plant Assessment Process

( Continued)

  • System Screening

- Utilization Of The Screening Attributes In Secy 93-331 And SRM (2/3/94), As Modified.

By Additional Experience

- Primarily A Systems Based Review

- Review Results List Of Systems Require That Further Evaluation

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Integrated Plant Assessment Process (Continued)

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  • System Evaluation

- Identify Equipment Types

- Concentrate On Long-Lived, Passive Structures And Components

- Maximize Utilization Of The Maintenance Rule And Other Existing Programs

- Utilization Of Renewal Programs As Necessary

  • PT194.15

Integrated Plant Assessment Process (Continued)

  • lime - Limited Analyses

- Applicable To SSCs Whose Safety Was Premised On Explicit lime Limited Analyses

- CLB Wi 11 Be Reviewed To Identify Any Applicable Analyses PT11M.18

Project Topical Reports (Surry PLEX Program) Addressing The Following Major Components And Structures Which Are Being Reviewed:

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Reactor Coolant Isolation Valves Control Rod Drive Mechanisms (CROM)

Neutron Shield Tank Pressurizer Steam Generators NSSS Supports Accumulators Primary Sh_ield Penetration Coolers Primary Shield Water Wal Is Main Turbine Turbine Pedestal Station Transfonners Containment Containment Internal Structures*

Containment *penetrations Containment Subsurface Drain Cable In Containment Structures (Other Than Containment)

Intake/ Discharge Structure Intake/ Discharge Canal Spent Fuel Pool / Transfer Canal Major Concrete Embedments Major Concrete Anchors

  • Control Room And l&C Major Piping (Other Than RCS)

Cable Outside Containment Emergency Diesel Generators Main Generator PT1114.17

Examples Of Existing Programs/ Activities That Assure SSC' Function I

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  • Program Attributes

~ Testing

-- Monitoring

- Inspection

- Maintenance

- Operability Status Determination

. - Refurbishment/Replacement/Repair PT11M.18

Examples O_f Existing Programs/ Activities That Assure SSC Function (Continued)

Programs/ Activities.

- Surry And North Anna Service Water Upgrade (GL-89-13)

- Reactor Vessel Surveillance And Integrity (GL-92-01)

- Flow Accelerated Corrosion (G-89-08)

- lnservice Inspection And Testing (ASME Section XI)

Technical Specifications Surveillance (1 OCFR 50.36).

- Containment Leak Rate Testing (Appendix J)

Fire Protection (Appendix R).

- Equipment Environmental Qualification (1 OCFR 50.49)

- Reliability Centered Maintenance Steam Generator Maintenance Agreement

- Maintenance Rule Implementation (1 OCFR 50.65)

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  • Safety And Relief Valve (I NPO Good Practice)

- (heck Valve (INPO Good Practice)

- Electrical Breaker (NRCB 88-01 And 88-10)

- Predictive Analysis And Maintenance

- Preventive Maintenance

. - Emergency Diesel Generator Reliability (GSI B-56)

. Nuclear Plant Chemistry (I NPO Good Practice)

PT11M.111

Phase 1 IPA Results

  • Reviewed 3 Systems

-AFW

- IA

- DC Power

  • Reviewed Various Generic Equipment Types (Raceways, Tanks, Cable, Structures, Supports)
  • Tested Methodology and Reporting Systems
  • No New Technical Issues Found
  • Some,Additional Monitoring (Cable, Concrete)
  • Finalizing Cost And Schedule Estimates*

i PT11M.20

Equipment Data System E. S. Grecheck Manager Inservice Inspection / Nondestructive Examination And Engineering Programs PT11M.21

Equipment Data System

  • Passport Is A Mainframe Based Relationa~

Database Application

  • PassportConsists Of Four (4) Main Modules

- EDS - Equipment Data System

- BOM - Bill Of Material System

- DMIS--Document Management Information System

- PWCS - Plant Work Control System

  • Engineering Responsible For EDS And BOM I

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I Equipment Data System EDS BOM PWCS DMIS Data Is Shared Seemlessly Between Applications I

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PT194.23*381

Equipment Data System

  • Owned And Maintained By Engineering
  • Contains Over 256,000 Mark Number Records
  • Controls The 37 Character Mark Number ID Number For Plant Components
  • Provides Descriptive And Programmatic Data For Plant Components..
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  • Flexible Parameter Sets Al low Sets Of Data Attributes To Be Applied To Groups Of Components
  • Accessible For Browse Without Computer Sign On PT11M.24

Q-List I

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  • A Subset Of EDS Records Made Up Of Components Classified As Either Safety Related (SR) Or Non Safety With Special Quality Or Regulatory Significance (NSQ).
  • Present Q-List Contains

- 88,200 SR Components

- 34,656 NSQ Components

  • Q-List Is Based On Functional Analysis Of Components.
  • Function* Will Form The Basis For Graded Q-List.

PT1114.25

Q-List Function Based Approach

  • The*Q-List Is Based On A Deterministic Review Of
  • component Functions.
  • Functions Are Derived From A Review Of UFSAR, SDBDs, Drawings, Training Modules, And Other Available Information.
  • Functions*Are Categorized Into Codes Defined In Engineering Standard. (SR, NSQ, Or NSR)

PT1114.28

Physical Verification Walkdowns I

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  • Performed Under Configuration Management Project
  • Took Over 4 Years
  • Completed In Early 1993
  • Components Categorized Into Functional Parts.

For Example:

  • - An MOV Can Have Up To 5 Mark Numbers

- Motor

- Operator

- Valve

- Open Limit Switch

- Closed Limit Switch

  • One Reason Q-List Is Large Is The Depth To Which Components Are Now Mark Numbered.

PT1114.27

Q-List Enhancement

  • Consists Of Three Phases Phase 1

- Classify All Components Added By The Physical Verification Project

- Involved 168,000 Components

- Completed 10/93 Phase 2

- Provide Detailed Functional Basis for Safety Classification For Al I SR Components Identified In Phase I

- Involves Approximately 48,000 Components

- Scheduled To Be Completed 12/31/94 Phase 3

- Provide Detailed Functional Basis for Al I NSQ Components Identified In Phase 1 And Confirm Functional Basis For Remaining SR Components

- Ongoing I

1 PT194.2B

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Virginia Power

  • Nuclear Analysis Capability K L. Basehore Supervisor Nuclear Safety Analysis I

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PT1114.211

Advantages Of In-House Capability For Virginia Power I

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  • Development Of Analysis Expertise

- Better Understanding Of True Impact Of Plant Issues

  • Better Plant Operational Flexibility

- Use Of Plant Specific Analyses Creates Margin For Plant Optimization

  • Efficiency Through Centralization Because Of Four Similar Nuclear Units

- Engineers With Cognizance Of Multiple Units

- Rapid Use Of Resources Into Active Analysis Area

  • . Consistency

- Implementation Of Methods, Assumptions And Setpoints

- Resolution Of Generic Issues

. PT11M.30

Nuclear Analysis*

And Fuel Mission 11Cradle To Grave" Responsibility For Nuclear FuerAnd Components

- Nuclear Fuel Management And Optimization Based On e

Outage Schedules Nuclear Fuel And Component Mechanical Design Specification Surveillance Of Fuel Fabrication, New Fuel Receipt Inspection Reload Core Design And Safety Analysis-Site Reactor Engineering, Startup Physics Testing And Core Follow Irradiated Fuel Inspection And Evaluation Long Term Disposition Of Spent Fuel, Including ISFSI

  • Reactor Engineering Function
  • . Reactor Vessel Survei I lance Program
  • Offsite Dose Analysis Evaluation
  • UfSAR Safety Reanalysis PT11M.31

NRC Approved Tools I~

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Codes/Models

- PDQ-7 (Discrete)

- PDQ-7 (Coarse Mesh)

- FLAME (3-D)

- NOMAD (1-D)

- RETRAN (System)

- COBRA (Core Thermal-Hydraulics)

Generalized Methodology

- Nuclear Uncertainty Factors

- Reload Evaluation Methodology Specialized Applications

- Relaxed Power Distribution Control

- Rod Ejection Analysis Methodology

- COBRM1\\IRB-1 Correlation

- Statistical DN B PT1114.32

Historical Examples Of Nuclear Analysis Capability I

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  • Reload Core Design And Safety Analysis

- Responsible For Full Scope On SPS And NAPS Since 1984 (e.g. NAPS 1 Cycle 5, Currently Designing NAPS 1 Cycle 11)

- Cycle Redesign Or Reevaluation When Required To Address:

Removal Of Leaking Fuel At End Of Cycle Extension Of Cycle Operation Increased Steam Generator Tube Plugging I

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PT1D4.33

Historical Examples Of Nuclear Analysis *Capability I

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  • Changes To Technical Specification Operating Limits

- SPS BIT Removal NAPS BIT/BAT Boron Concentration Decrease Relaxed Power Distribution Control For NAPS (Topical Report)

NAPS Revised EOC MTC Measurement Strategy (Recommended GL 93-05)

Revised FQ And K(z) At NAPS And SPS5everal limes Revised FdH And OT~T At SPS (Statistical DNB Topical Report)

RWST Boron Concentration Increase For NAPS And SPS Revised Min,imum RCS Flowrates, NAPS Revised Heatup/Cooldown Curves And LTOPS Setpoints Reduced RCS Pressure For PSRV Leakage Issue At SPS PT11M.34

Historical Examples Of Nuclear Analysis Capability ( Continued)

  • Changes To Other Operating Limits

- Temperature Coastdown At End Of Cycle For SPS And NAPS

- Steam Generator Plugging Limits (LOCA Capability Since 1981)

- Changes In Steam Generator Mass Inventory

  • Physical Plant And. Equipment Changes

- North Anna Steam Generator Replacement

- RDT Bypass Elimination

- North Anna Core Uprating (Portion Of Safety Analysis)

- Surry Core Uprating On-Progress)

- Fuel Product Upgrades, NAPS And SPS

PT1114.35

Historical Examples Of Nuclear Analysis Capability ( Continued)

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  • Emergency Procedure Basis

- Technical Basis For SPS And NAPS Implementation Of WOGERG

- Technical Basis For Loss Of Shutdown Cooling Abnormal Procedure

- Technical Basis For Shutdown LOCA Abnormal Procedure

- WOG Severe Accident Management Guidelines Task Team

- AFW Crossconnect Technical Specification Change

- SPS IPE (SER Received 12/93)

- NAPS IPE (Submitted On 12/92)

- NAPS IPEEE (To Be Submitted 6/94)

- SPS IPEEE (In Progress) 1 OCFRS0.65 Maintenance Rule - Risk Significant SSC Ranking I

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Surry Core Uprating Status K L. Basehore

  • Supervisor Nuclear Safety Analysis PT1114.37 I,

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General Approach

  • Uprate To Original 11Stretch 11 Power Level Of 2546 MWt And 856 MWe - (Current Power Level Is 2441 MWt And 822 MWe)
  • Analysis And Evaluations To Support Core Uprate Project Have Been Ongoing.
  • Current Effort Involves Validating And Updating Previous Results To Reflect Current Plant Condit~ons
    • Preliminary Conclusions Demonstrate Acceptability Of Plant Systems And Components For Uprated Operation PT194.38

Original vs. Uprated RCS Design Parameters Original U12rated Reactor Coolant System Core Power, MWt 2441 2546 Thermal Design Flow, GPM 265500 265500 Mass Flow, Million LBtvVHr 100.7

  • 101.1

., Pressure, PSIA 2250*

2250 Temperature, °F Vessel Outlet 605.6 605.6

.. Vessel Average 574.3 573.0 Steam Generator Outlet 543.0 540.2 Steam Generator Steam Temperature, °F 516.0 516.0 Steam Pressure (Min. Guaranteed), PSIA 785.0 785.0 Total Steam Flow, Million LB!vVHr*

10.66 11.23 Feedwater Temperature, °F

  • 437.7
  • 443.0 Zero Load Temperature, °F 547.0 547.0 PT11M.311

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Surry Core Uprate Project Scope NSSS And BOP Evaluated For

- Consequences Of Accidents Postulated In UFSAR

- Capability Of Systems And Equipment To Meet Design Bases Specified In UFSAR I

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- Operating Limits And Conditions Contained i

In Technical Specifications Determine Need For Plant Modification PT111UO

Safety Analysis Evaluation

  • NSSS Accident Analysis

- The Majority Of Safety Analysis Are Sen~itive To Power Level And Have Been Reanalyzed By Virginia Power.

- Where Appropriate New Methods Have Been Used, e.g.

Large Break LOCAAnd Statistical DNB.

  • Accident Radiological Consequence Evaluations

- Offsite And Control Room Dose Calculations Have Been Completely Reanalyzed By Virginia Power, Using Assumptions Consistent With NUREG-0800 Guidelines.

  • Containment Integrity Analyses

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- Operation At Uprated Conditions With Current Tech Spec Limits Requires Improvement In Analysis Margin.

  • - Margin Available Through Model Inputs, Analytical Improvements And Better Interface Between Mass And
  • Energy And Containment Response Assumptions.*

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PTIIM.41

Summary Of Results To Date

  • NSSS Systems And Components Are Acceptable And In.

Compliance With Applicable Codes And Standards

  • BOP Systems Are Acceptable With Minor Exceptions To Be.

Addressed Prior To Uprated Operation

  • NSSS Accident Reanalysis And Radiological Consequence Evaluations Meet Appropriate Design Criteria And Support Operation At Uprated Conditions
  • Containment Integrity Analyses In Progress Are Projected To *
  • Show Increase In Margin To Key Containment Design Limits PT1D4.42

Plant Modifications*

  • Completed Modifications

- Steam Generator Moisture Carryover Modifications On 1 A,

. 1 B,* 1 C, And 2A

  • Planned Modifications

- Fifth Point Feed Heater Drain Nozzle Reinforcement During Unit 2 1995

  • Potential Modifications

- Steam Generator Moisture Carryover Modifications On 2B And2C

- Testing May Show That Modifications Not Necessary

  • Protection And Control Setpoint Revisions After License Amendment Approval PT11M.43

Core Uprate Schedule

  • Completion Of Validation And Reanal-ysis By May 1994
  • Submittal Of License Amendment Planned
  • July 1994
  • Implementation At Power Is Planned (As Was.

Done For North Anna)

PT11M.4-4 I

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Design Basis Documentation Project Update.

D.L. Benson.

Manager Nuclear Engineering PTIIM.46 I

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Agenda. -.Design Basis Documentation Activities

  • Document Assessment
  • Design Basis Documents
  • Identification, Review And Resolution Of Open Items PT1114.48 I

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Document Assessment 1

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  • Document Assessment Program Ensures That All Existing Design Information Is Utilized In DBD Development And Is Available To Support Engineering Activities.
  • Document Assessment Process Involves Collection, Review, Coding, And Image Scanning Of The Following Documents:

Licensing Correspondence NE Descriptions Calculations Technic~I Reports Project Correspondence Plant Life Extension Reports Design Changes Training Modules Engineering Work Requests UFSAR Preop And Start-up Tests Technical Specifications Setpoint Changes NSSS Information A/E Equipment Specs

  • The Result Is A Computer Based Document Index Allowing Sorting By Multiple Criteria And Retrieval Of Document Images.
  • Status: Over 2,500,000 Pages, 120,000 Documents In The System Index And Images Will Continue To Be Updated Annually.

PT11M.47

Design Basis Documentation

  • DBD Program Approved And Initiated In 1989
  • Scope

- Prepare System DBDs For Major Systems At Each Station (Approximately 57 Per Station)

- Prepare Plant DBD For Each Station

  • Schedule

- Seven Year Program, 1989 Through 1996

- Approximately 7 System DBDs Per Site Per Year

  • Current Status

- 32 System DBDs Per Station Either Issued Or In Draft Complete Stage

- 9 Draft System DBDs Per Site To Be Completed By Year End *

- Approximately 16 Remaining Per Site

- Plant DBD: "Accident.Analysis DBDs" Will Be Completed For Each Site In 1994

  • On Schedule For Completion 1*n 1996 PT111U8 I

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Identification, Review And Resolution Of Open Items Identification o "Open Items" Alert The DBD User To Any Qualifications Relative To The Data Provided.

  • "Open Items" Are Summarized In the Back Of The DBD And Include:

Missing Or Incomplete Source Documentation Or Referencing Of A Secondary Reference Conflicting Information Or Data In Source Or Reference Documents Errors Or Discrepancies In Calculations, Reports, Etc.

Missing Or Incomplete Calculations Linkages To Other DBDs Not Yet Completed Other Matters Judged To Require Additional Review Or Action

. Review And Resolution

  • Preparers Are Required To Notify Virginia Power Immediately Of Any Safety Or Operability Concerns
  • "Open Items" Are Reviewed by A Multi-Discipline Team Following Completion Of The Draft DBD. Items Are Grouped Into Three Categories:

Immediate Resolution Intermediate Resolution Resolve If Needed PT11M.50

Actions Taken Or In Progress Related To Open Items l

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,t Concerns Of Inspection Report 92-32

  • Review Of Open Items For Safety Significance
  • Size Of Backlog

Response

  • All Open Items Existing At The Time Of Inspection 92-32 Have

'Been Re-Reviewed By A Multi-Discipline Team. No Safety Issues Were Identified. Review Resu Its Have Been Presented ToSNSOC's.

  • One Third Of Open Items Existing At The Time Of IR 92-32 Have Been Resolved.
  • All Remaining Open Items Now Have A Written Basis For Deferral, A Designated Responsible Party And A Due Date For Resolution.
  • Minutes Of Future "Open ltem 11 Review Meetings Will More Fully Document The Bases For Deferral.
  • Goals For Resolving Remaining Open Items Are Tracked In Our Backlog Reports.
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Summary-Design Basis Documentation

  • Document Assessment Project Has Provided A Comprehensive Design Information Base From Which To Prepare DBDs. The Database And Images Will Be Maintained to Support Engineering Activities.

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  • DBDs Are Thorough, Consistent, And Are Designed To Be Maintained And Augmented. DBD Development Is On Schedule For Completion In 1996.
  • Open Items Are Properly Assessed For Reportability And Unreviewed Safety Questions. Documentation Of The Basis For Deferring Open Items Has Been Improved..

PT194.62

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