ML20035E899

Summary of 930402 Meeting W/Util in Region 1 Ofc to Discuss Plant Ipe.List of Attendees & Copy of Licensee Slides Encl

ML20035E899
Person / Time
Site:	Brunswick
Issue date:	04/09/1993
From:	Merschoff E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	Watson R CAROLINA POWER & LIGHT CO.
References
GL-88-20, NUDOCS 9304200041
Download: ML20035E899 (35)

Text

{{#Wiki_filter:______ _ [i ( ADR 9 1993 [ Docket Nos. 50-325, 50-324 License Nos. DPR-71, DPR-62 Carolina Power and Light Company ATTN: Mr. R. A. Watson Senior Vice President Nuclear Generation P. O. Box 1551 Raleigh, NC 27602 Gentlemen:

SUBJECT:

MEETING

SUMMARY

- BRUNSWICK This refers to the management meeting conducted at NRC's request in the Region II Office on April 2, 1993.

The purpose of the meeting was to discuss Brunswick's Individual Plant Examination (IPE). A list of att. iees and a copy of your slides are enclosed. 4 It is our opinion that this meeting was beneficial, in that it'provided a better understanding of the IPE results and its application at Brunswick. In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, a copy of this letter and its enclo-sures will be placed in the NRC Public Document Room. Should you have any questions concerning this matter, please contact us. Sincerely, Original signeo by: Ellis W. Merschotf Ellis W. Merschoff, Director Division of Reactor Projects

Enclosures:

1. List of Attendees 2. Licensee Slides cc w/encls: R. A. Anderson Vice President Brunswick Nuclear Project P. O. Box 10429 Southport, NC 28461 (cc w/ encl cont'd - See page 2) 9304200041 930409 PDR ADOCK 05000324 I P PDR ZE0/

Carolina Power and Light Company 2 ApR g [ggg (cc w/ encl cont'd) M. Brown Plant Manager Unit 1 Brunswick Steam Electric Plant P. O. Box 10429 Southport, NC 28461 C. Warren Plant Manager Unit 2 Brunswick Steam Electric Plant P. O. Box 10429 Southport, NC 28461 Mr. Mark S. Calvert Associate General Counsel Brunswick Steam Electric Plant P. O. Box 10429 Southport, NC 28461 Kelly Holden Board of Commissioners P. O. Box 249 Bolivia, NC 28422 Dayne H. Brown, Director Division of Radiation Protection N. C. Department of Environment, Health & Natural Resources P. O. Box 27687 Raleigh, NC 27611-7687 H. A. Cole Special Deputy Attorney General State of North Carolina P. O. Box 629 Raleig5, NC 27602 Robert P. Gruber Executive Director Public Staff - NCUC P. O. Box 29520 Raleigh, NC 27626-0520 Ms. Gayle B. Nichols Staff Counsel SC Public Service Commission P. O. Box 11649 Columbia, SC 29211 bcc w/encls: (See page 3)

APR 9 1993 Carolina Power and Light Company 3 bcc w/encls: Document Control Desk H. Christensen, RII P. Milano, NRR NRC Resident Inspector U.S. Nuclear Regulatory Commission Star Route I, Box 208 Southport, NC 28461 RI,DRP RII:I) RII: [ RII:DR \\ DVerrel\\e N,ll:tj HChr stensen i JJohRson K 04/7/93 04/A/93 04/ '}/93 04/ /93

ENCLOSURE I Carolina Power & Liaht R. Watson, Senior Vice President, Nuclear Generation J. Cowan, Manager, Regulatory Affairs G. Attarian, Electrical Engineering Manager, Nuclear Engineering Department (NED) G. Miller, Risk Assessment, NED R. Oliver, Risk Assessment, NED F. Emerson, Nuclear Licensing Nuclear Regulatory Commission S. Ebneter, Regional Administrator, Region II (RII) L. Reyes, Deputy Regional Administrator, RII J. Johnson, Deputy Director, Division of Reactor Projects (DRP), RII A. Gibson, Director, Division of Reactor Safety (DRS), RII J. Mitchell, Acting Director, Project Directorate PD Il-1, Office of Nuclear Reactor Regulation H. Christensen, Chief, Projects Section 1A, DRP, RII R. Crlenjak, Chief, Operational Programs Section (OPS), DRS, RII J. Shackelford, Reactor Engineer, OPS, DRS, RII G. Harris, Project Engineer, DRP, RII

ENCLOSURE 2 CAROLINA POWER & LIGHT COMPANY BRUNSWICK PLANT IPE RESULTS REGION 11/ CP&L MEETING ATLANTA, GEORGIA APRIL 2, 1993

INTRODUCTION o Responding to Region 11 request for information on Brunswick Plant Individual Plant Examination (IPE) results o Will share information on IPE results and what they mean "Use and usefulness" of PRA at BSEP 2

SUMMARY

o Brunswick Plant IPE performed in accordance with Generic l.etter 88-20 o Resulting core damage frequency of 2.7 E-5 in middle of range for comparable plants o Principal insight that loss of offsite power events are important has been addressed through actions taken since submittal of IPE o Process in place for handling IPE results and other severe accident issues in accordance with NRC and NUMARC guidance o IPE is being maintained and applied for the Brunswick Plant e 3

AGENDA NRC REGION 11/ CP&L MEETING ON BSEP IPE APRIL 2,1993 Time er33 Speaker r 1500 Introduction / Summary R. A. Watson (Senior VP NGG) 1505 IPE, IPEEE background / requirements F. Emerson (NLS) 1515 BSEP IPE findings G. Miller (NED/PRA) Level 1/ 2 results Comparison with other Region 11 plants insights 1530 AC power system / Diesel generator capabilities G. Attarian (NED/RESS) 1540 S<tvere accident issue management process Emerson Handling of results Project team conclusions Management concurrence Submittal 1550 What makes PRA a useful tool Emerson Risk importance measure 1555 Applications Miller 3 year plan prioritization DBD Others 1605 Potential future uses at CP&L Miller PRA Update 1610 Questions / discussion All 4

IPE / IPEEE BACKGROUND o BSEP submitted a PRA in 1988 which included Level 1 Limited Level 2 External events o Generic Letter 88-20 ...the Commission concluded... that existing plants pose no undue risk to the public health and safety and that there is no present basis for immediate action on generic rulemaking or other regulatory requirements for these plants." "However, the Commission recognizes..that syste-matic evaluations are beneficial in identifying plant - specific vulnerabilities to severe accidents that could be fixed with low - cost improvements." Purpose is for utility to gain appreciation for SA behavior understand most likely SA sequences gain quantitative understanding of core damage and containment release probabilities if necessary, reduce core damage and containment release probabilities 5

IPE / IPEEE BACKGROUND o GL 88 - 20, Supplement 1 Identified Mark 1 - containment performance improvements to be included in IPE o GL 88 - 20, Supplement 2 Identified accident management strategies for possible inclusion in the IPE process o GL 88 - 20, Supplement 3 Identified containment performance improvements for other than Mark I containments o GL 88 - 20, Supplement 4 Requested IPE for External Events (IPEEE): Seismic events Internal fires High winds and tornadoes External floods Transportation and nearby facility accidents 6

OVERVIEW OF BRUNSWICK IPE RESULTS o Level 1 IPE identifies accident sequences that lead to core damage o Level 2 IPE takes the Level 1 core damage sequences and considers vessel melt-through and interactions within the containment to the point of containment failure o The most visible product of the Level 1 IPE is the Core Damage Frequency, (CDF) o The most important product is the insights on important contributors to risk o The CDF for Brunswick IPE was 2.7 E-5 per year per unit o Brunswick CDF is within the range of values for similar BWRs 7 iii i i

OVERVIEW OF BRUNSWICK IPE RESULTS l 4 Comparison To Other IPEs ) Plant -CDF (/vr) l FitzPatrick 1.9E-6 l l Peach Bottom 5.5 E-6 Millstone 1 1.1 E-5 i t Hatch 2.1 E-5 i Brunswick 2.7 E-5 Browns Ferry 2 4.8 E-5 I WNP-2 5.4 E-5 ~ I = Pilgrim 5.8E-5 i i Differences Between Reported CDF Values i o AC Power Supply 1 o Cooling Water l o Plant Operating History i 1 o Internal Flooding l 8 i i ... _ _.. _ - -. _ -. ~.. - _. - -. _. _,... - -.. _ - ~. _ -... _ _ _ _. -.., -. - ... _ _ =. _, _

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i d 0 KEY ACCIDENT SEQUENCE o The dominant accident sequence is a station blackout sequence (Iow probability, high % contributor) o Sequence of Events: Loss of offsite power and failure of diesel generators (t = 0) High pressure injection systems (HPCI, RCIC) operate until station batteries depleted (t = 2 hr) Depletion of the batteries means remote operation of the switchyard breakers no longer possible Recovery of either the diesel generators or the l ) inter - unit emergency bus cross - tie were j assumed to be unlikely because of the short time available l After HPCI and RCIC failure, water level in the core decreases a Core uncovery begins at t = 4 hr The vessel fails at t = 12 hr Containment failure (leakage) occurs at t = 22 hr 10 --,.,-,,-,-----.,,-e,.~,---,,--~,,m-- -,w,,,e,,---ew,-ww--- ,,,,-n-----,,ame-m,,r-,nne.--.-rw-,e,,, ewe,.,,-,-rs,-,.w, e --v,e -nm,-,,we ,,w,ovs.-o----

l INSIGHTS FROM THE IPE o The calculated Core Damage Frequency of 2.7 E-5 per year was within the expected range of BWRs o The availability of AC power and decay heat removal j is most important o Almost all containment failures are benign and self regulating leak before break failures (all releases below WASH - 1400 values) Drywell head flange leak at 120 psig Torus / drywell failures at 257 / 322 psig o Control Rod Drive injection from Condensate Storage Tank could keep core covered for several days Injection piping and pumps at lower elevations are not affected by principal failure mode of drywell head leakage o Liner melt - through failure mode at BSEP does not lead to large release due to reinforced concrete design o BSEP containment therefore offers substantial benefit 11

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CURRENT PRA MODEL STATUS o The PRA model is currently being updated: Plant modifications and procedure changes l l l Hardened Wetwell Vent Emergency Bus Cross - tie Station Blackout Procedure Model Enhancements CRD injection for Decay Heat Removal Revised operating data 1 Recent loss of AC power event Other operating data changes o Preliminary estimate of revised CDF: 1.3 E-5/yr l l 14 l l l l

AC POWER SYSTEM / DIESEL GENERATOR CAPABILITIES o Electrical Configuration o Diesel Generator Description / Ratings t o Diesel Generator Design Basis o Diesel Generator Operating Scenarios i G 15 --r-w ,--wv w v- -g y.+ pw-ee ,irp-w -rw-9 y g g

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BSEP AC EMERGENCY POWER SYSTEM i 1D 1C 2D 2C h f i u i I _l E4_ 1 E1_ 1 _l E2_ l E31 ) ) )') ) ') ) ) ) )') )) ) O O b O EDG1 EDG2 EDG3 EDG4 $U DU ) ) ) ) E7 E8 ES E6_ ) ) ) ) ) ) ) MCC's MCC's MCC's MCC'S l 17 - e_

i 4 DIESEL GENERATOR DESCRIPTION / RATINGS I i Engine Manufacturer Nordberg Manufocturing Company 3 i Engine Model Number FS-1316-HSC i Engine Type V-16' / 4 cycle / 514 rpm ] (turbocharged) l Engine Governor Woodward Model # EGB-35P/LS Generator Manufacturer General Electric Company Generator Type Salient Pole / 514 rpm / Self - Cooled Generator Voltage 4160 vac / 3 - Phase / 60 hz Diesel Generator Ratings l Continuous 3500 KW Overload 3850 KW (2000 hours / year) l 18 i

1 DIESEL GENERATOR DESIGN BASIS o Diesel generator capacity sufficient to supply Emergency Safety Features under DBA on one unit, and l for safe shutdown on the non - event unit. o Assumptions Loss of offsite power on both units I l Failure of one diesel generator l 19 l m -r-,-m-- -wv,.., .-,-,ww-- ---,re-m.e+=---++-m-.-++-e ww.y-,y--w---

DIESEL GENERATOR OPERATING SCENARIOS LOOP (4 Diesel Generators) Load (KW) 3472 Rating (KW) 3500 i l i LOOP (3 Diesel Generators) Load (KW) 3097 Rating (KW) 3500 LOOP + Accident (4 Diesel Generators) i l I Load (KW) 2554 Rating (KW) 3500 l l l I LOOP + Accident (3 Diesel Generators) Load (KW) 3428 Rating (KW) 3500 l l Station Blackout l Load (KW) 3830 Rating (KW) 3850 20 L

OBJECTIVES OF CP&L SEVERE ACCIDENT ISSUES MANAGEMENT PLAN o Approved by NGG management in July,1991 o Objectives l Assure safety to maximum extent l l Utilize Company -resources prudently j Minimize impact on plant operations Provide for consistent response to issues among plants Utilize Company investment in IPEs to select and schedule projects with high safety value l I l o Used to address principal NRC severe accident closure elements IPE results (1992 - 1993) IPEEE results (1994 - 1995) Severe accident management (1993 - 1995) 21

l l l ~ i r PROJECT TEAM CONCEPT i t i i o One project manager for the Company i i l o One project team for each site working with project l manager l l l I t j o Project team includes representatives from 1 Plant Operations { Plant Tech Support i 1 Plant Training-PRA I Corporate Engineering l I l Licensing Emergency Preparedness Legal I 22 l ...,-_.,,,,..-...,...,m..-,, ,,m_,m__,,.,_,. _., _.,. -,, _ _,.,_,._.,.. _.....,.

HANDLING OF IPE RESULTS I o Developed criteria for addressing IPE results based 'on NUMARC closure guidelines I l Qualitative cost / benefit considerations Effect on plant operations l l o Assessed IPE results against these guidelines l l l o Team recommendations reviewed by plant management o Final recommendations approved by nuclear senior management i l I 23

BRUNSWICK PLANT IPE RECOMMENDATIONS o No additional enhancements were recommended to address IPE results (CDF = 2.7 E-5 / yr) o At the time of the submittal, previously planned improvements estimated to provide 70 % CDF reduction 5th diesel E - bus crosstie improvement Hardened wetwell vent l o Since submittal of results to the NRC in August 1992, completed improvements estimated to provide l 50 % CDF reduction SBO procedure revision l E - bus crosstie improvement Hardened wetwell vent i More realistic CRD injection modelling l o 5th diesel estimated to add a further 25 % CDF reduction; therefore, may not be cost beneficial from an IPE risk reduction standpoint l i 24

WHAT MAKES PRA A USEFUL TOOL _ CHARACTERISTICS i o it considers the systems - that can be beneficial in preventing core damage l 0 It reflects actual conditions, system / component interdependencies, and operator use of plant procedures i l event trees model system interdependencies and l operator actions fault trees model individual syt, tem operation, including operator action human reliability analysis reflects the actual time, i procedural, and training ' environment impacting l operator responses l o Actual plant data used for initiator (scram) frequencies equipment failure probabilities for important components (DG, HPCl, RCIC, SWS, RHR) system out of service time for most systems 25

WHAT MAKES PRA A USEFUL TOOL OUTPUT o it provides importance rankings for individual component and operator actions considering probability of failure system and / or component redundancy o it can be used to develop system importance rankings 1 26 +

WHAT MAKES PRA A USEFUL TOOL i APPLICATION o its greatest value is as an objective tool for allocating resources to the projects or activities with the greatest benefit in protecting the core or containment providing a useful perspective on the safety benefit of taking or not taking certain actions BUT o it must be used in conjunction with other important considerations such as I ^ ALARA l personnel safety i licensing / design basis cost l plant operations l 1 27 4 I l

PRA APPLICATIONS i l o 3 Year Plan Prioritization 1 A systematic method --of prioritizing plant projects PRA importance ranking of systems is the basis for the Nuclear Safety Scaling Factor i o Design Basis Document Verifir Son l PRA used to assess the safety significance of l l design basis discrepancies identified in the DBD l process ) l l o Plant Support Risk - based assessments to support operations and outage activities Prioritization of work items Operator training i 28 y ,er-r--- g -,y-y y y,m+,, --p-- wyg. g s-m-y-r-.E a we e &w g-tuw Tur t-emg-7 ey mey' 'M - e tMy-wmy-W l PRA APPLICATIONS I o IPE for External Events June 1995 completion for

Fire, Seismic, and Others o

Maintenance Rule Identify risk - significant components f i Develop risk - significant criteria l o Severe Accident Management Incorporate industry guidelines into plant - specific actions i 29 w-a T----- -e y-wsi ---y-p,,,,.,y .,,,,y w, y,,,, -,,py y w--gy+%-+,- ym-+m-vep-- --wyg-,y,y-y3

( PRA APPLICATIONS o Possible PRA Applications Establish Plant Risk Goals MOV Testing (GL 89-10) Optimization Simulator Training Support Shutdown Risk Emergent issues l l l 1 30 i I

PRA UPDATE o The PRA model is updated during each refueling 1 outage Reflect design and procedure changes Update operating data ) Modeling improvements o An updated importance Analysis is performed l Basis for prioritization 31 W F--*T e*aa we-

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