ML070510497

G20060793 - John D. Runkle Ltr. Re 2.206 Petition - Shearon Harris Fire Safety

ML070510497
Person / Time
Site:	Harris
Issue date:	02/08/2007
From:	Runkle J D Conservation Council of North Carolina
To:	Regner L M Plant Licensing Branch III-2
References
2.206, G20060793
Download: ML070510497 (6)

Text

JOHN D. RUNKLE ATTORNEY AT LAW POST OFFICE BOX 3793 CHAPEL HILL, N.C. 27515-379 919-942-0600 3 February 8, 2007 EDO DEDMRS DEDR DEDIA AO cýc1 Lisa M. Regner Project Manager Division of Operating Reactor Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Re: 2.206 Petition -Shearon Harris Fire Safety

Dear Ms. Regner:

As part of the review of our 2.206 Petition on the fire safety issues at the Shearon Harris Nuclear Plant, I would like to bring to the attention of the review panel another relevant document that may not have been available to them for their review. The document is the Response by CP&L, now Progress Energy, to Generic Letter 88-20 Supplement 4 -Individual Plant Examination for External Events (IPEEE), dated June 30, 1995. I am attaching the transmittal letter and the relevant page that addressed internal plant fires. The entire 430-page document is available in the NRC's Public Document Room under Accession no. 9507060075, but not in ADAMS. Page 2-6 of this IPEEE states: 2.2.2 Internal Plant Fires The object of this task is to estimate the contribution of accident sequences induced by in-plant fires to overall core damage frequency.

The fire evaluation was performed on the basis of fire areas, which are plant locations completely enclosed by rated fire barriers.

The fire area boundaries were assumed to be effective in preventing a fire from spreading from the originating area to another area based on the implementation of a satisfactory fire barrier surveillance and maintenance program.CP&L erroneously assumes that its fire barriers were 100% effective in preventing a fire from spreading from room to room in its calculation that the overall fire hazard was very low. When evidence subsequently surfaced that the fire barriers were far less than 100% effective, CP&L used the erroneous results showing fire hazards to be very low to dismiss the significant safety problems at the Harris Plant.1~itpLzkei 6bo~c~o~Of Harris 2.206 Petition, page 2 -Please see that this letter and accompanying attachment are provided to the panel reviewing the Petition.Thank you for your attention to this matter.Sincerely, John D. Runkle For Petitioners Enc.cc. Jim Warren, NCWARN Paul Gunter, NIRS David Lochbaum, Union Concerned Scientists John H. O'Neill, Jr., Pillsbury Winthrop Shaw Pittman LLP for Progress Energy F0o &W 65 Yke hid NlpwM~ NK: 27562 ?*uc:6m JUN 3 0 t " File Number:'1HO-gO900 SERIAL: HNP-95-061 United States, Nuclear Regulatory Comumission ATTENTION:

Docmeff Control Desk Wahingon.

DC 20a55 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50240LCENSE NO. NPF-63 RESPONSE TO GENERIC LETTER 88-20 SUPPLEMENT 4 -INDIVIDUAL PLANT EXAMINATION FOR EXTERNAL EVENTS ([PEEE)Gentlemen:

.The pur'pose of this letter is to submit the results of the Individual Plant Examination for Externa Events for Carolina Power and Light Company's (CP&L) Searon Harris Nuclear Power Plant W(SHNPP)as con. mtted inour nerofc 15, 1 "2 (we eclosure).

The IPEEE was2comp0ed in accordance with Generic Let er 88-20, Supplem t 4 and the methods outlined in NUREG-1407

-,(fth NRC's procedural and submittal guidance).

Evaluation of seismic risk was performed using tho Seismic Margins Assessment methodology developed by EPRI. Evaluation of fire risk was perfoXMed using A t EPRI Fire-induced Vulnerability Evaluation (FIVE) methodology combined with a traditional fire PRCA. Evaluation of risk from other external events (i-aluding high winds, wdernad flooding, and trasotton and nearby facility accidents) was performed by _2__onstI

1. -the puros of thi lets oue 1975 Srttanrd Review Plan (SRP) criteria for thse external foeEtse The enults of the sCismic n Poer indicate Com p re an no significant seismic Noucema Six minor modificationiarcew irs will be completed by the end of Refueling Outage (RFO) 7N Urenttly scheduled for spring 1997. Examples include restraining caf and cabinets to preclude potential impatinaction nd asachin two cabin together, also to Eprclude pote al impact/ittera son.The fore IPEEE results indicate -Icore damage fru uency Evauaton (FIVE)imaetho.oIE-5 fomin the w signithatraditionafi re scenevios.

Per NUMARC/NEr 91-e , "Severe Accident ghhe Closure Guidelinesra it was not necessary to avndute modificationr or administrative dhanges to addrbss the scnmerios.

However. one Standard enhavPar.,n( t related to remote shutdown will be The res to vterfy the status of the pthatrth power operated relief valves after trnsfeS to the Auxiliary Control Panel and to r'equire isolmion in case of a failed open reliefval~e.

This procexlre isitwill be on rpics ior to stabc m p from RFO 6e cufenty scheduled to begin in September 199ri Using tn e NUMARCt(1 ) 91-04 snieeoines for NPEEE closure, CP&L expects to consider the=e rs during the deveroneept of plcrntspee ieh c Severe Accident Mote ms nt Guidance.Svi w e oed 1134 Now Hila Nf Tl 919 362-2s02 ed 919 362-209bg n

• Document Control Desk HNP-95-061

/ Page 2 As specified in NUIREG-1407.

no estimate of core damage frequency is required for other external events since our review showed that SIINPP complies with the 1975 SRP for these external events.This was determined by a re.iew of information available in the Final Safety Analysis Report (FSAR), by collecting supplemental information that might have changed since the last FSAR revision, and by prfonrming a confirmatory plant walkdown.In accordance with Generic Letter 88-20, the external cvents portion oft rSI A-45. "Shutdown Decay lieal Removal Requirements," are subsumed tithin the IPF-H- and are therefore considered resolved.

The Eastern United Stales Seismicity Issue and GCeneric Issue- 131 (seismically induced failure of tlux mapping transfer cart) are likewise considered resolved.

Als.o the Fire Risk Scoping Study Issues in NI IRE(i/CR.5088 were examined and addressed.

Finally. this IPI.EE addresses the revised "[i-sign Probable Maximum Precipitation" criteria (Generic Letter 89-221 and Hunicane Andrew lessons learned (Information Notice 93-53. Supp. 1).Questions regarding this matter may be referred to Mr. R. W. Prunly at (919) 362-2030, Sincerely.

RWP/rwp EInclosure W. R. Robinson.

having been first duly sworn, did depose and say that the intbrmation contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his infnonation are officers, employees, contractors, and agents of Carolina Power & Light Company.Notary(My commission expires: --4 --c: Mr. S. D. Ebntcer Mr. S. A. Elrod NOTAPY Mr. N. B Le PUBLC Shearon Harris Nuclear Power Plant Unit No. I Individual Plant Examination for External Events Submittal CAROLINA POWER & LIGHT COMPANY June 1995 950630 05000400 p POR In do case of active electrical and control equipment, it may nm be posible or corn effective to demonsM ity on the bets of achieved te level or by use of generic equipmmnt respone spectra (SQUG). The systems engineers are requind to evaluate the electrical ciruwits and operations procedures to assess the consequences and recovery wation for relay chatter, breaker trip, etc.2.2.2 Internal Plant firs The object of this task is to estimate the contribution of accidem sequences induced by in-plant fire to overall core damage frequency.

The aialysis consider, the likelihood of fire oecurren in each plant ar and its ubtequent impact on plant systems. Equipamnt damage reeking from the thermal effects of fitre (conductive, radiative and convective) are considered as well as the degradation of operation reliality.

Potential vulnerbilties raised in the Sandia FPS related to seismic/fire interactions, effects of suppressants on safety equipment and control system interactions are addressed through specifically tailored walkdowns, as defined in the EPRI FIVE methodology.

The models were developed in a systematic manner which enables the specific stengths a54 weaknesses of plant defenses against fire to be clearly identified.

The fire evaluation was performed on the basis of fire reas, which are plant location completely enclosed by rated fie barrien. The fire area boundaries were asumed to be effective in preventing a fire fom spreading fom the origimting ara to ameer area based on the implementation of a stisfactory fire barrier surveillance an mame'eance progmram.

The finr area bondaries recognized in this study ar identical to thou identified in the plato's Safe Shutdown Analysis (SSA) (CP&L. SSD). In some cases these fim arm were Arther subdivided dinto an for analysis purposes, for the more sigificant omparneu fire damage states within those compamnm s were defined that identified subse of the equipamnt within the comparmm as being damaged due to the fire.The analysis was condwud in thre main stages as follows: Stae I is a systematic qualitative and sMc ning aadysis of All plant fire arec/zones, foHowirg the methodology described in FIVE, Phase 1 aSd Phase 2, steps I and.2. The screening analysis was based largely on information already available in the plant's SSA a the IPE study. This resulted in the identification of fire arms and companmenuts in accordance with the FIVE methodology.

At this stage all equipment and cable in an area/compartment is assumed to be damaged. The damage was assessed qualitatively to determine if the effects wer signlfica that is, whether the fire would caus a plan shtAdown or trip, or lead to loss of ae shutdown equipment.

Areascompartmnts not screened out qualitatively were then subject to a detemiat.ion of their associated fire freqency (F 1) and conditional core damsage (Ps), given loss of all fuctiom which may be impacted by the re. If the reswulting fir induced core damage frequency (F, x Pz was less than IE-6 per year the mre/compalment was screened out.61V