ML20043F161

Provides Commission W/Status Rept & Staff Conclusions Re Mark I Containment Performance Improvement Program. Hardened Vent Capability Mods Expected to Be Implemented on All Mark I Facilities by Mar 1993

ML20043F161
Person / Time
Issue date:	06/07/1990
From:	Taylor J NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
To:
References
TASK-PII, TASK-SE GL-88-20, GL-89-16, SECY-90-206, NUDOCS 9006140208
Download: ML20043F161 (94)

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POLICY ISSUE June 7, 1990 SECY-90-206 For:

The Commissioners o

From:

James M. Taylor Executive Director for Operations

Subject:

MARK 1 CONTAINMENT PERFORMANCE IMPROVEMENT PROGRAM

Purpose:

To provide the Commission with a status report and the present staff-conclusions relative to the implementation of the subject t

program as outlined in the Commission's directives in the staff requirements memorandum, dated July 11, 1989.

Background:

In SECY-89-017, dated January 23, 1989, the staff presented its

l findings concerning the Mark I containment performance improvement

.i (CPI) program. On July 11, 1989,- the Commission responded to those findings by issuing a set-of directives to the staff. The y

status of the staff's earlier actions relative to the Commission's directives was summarized in SECY-90-023, dated January 17, 1990.

1 As stated in SECY-90-023.-on September 1, 1989, the staff issued i

Generic Letter 89-16 to inform the licensees of boiling water reactors (BWRs) with Mark 1 containments, that the Commission approved one of the staff's recommendations regarding hardening of the suppression pool vent path for generic implementation.

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The remainir.g staf f recommendations were to be incorporated in the licensee's Individual Plant Examination (IPE) activities and w

were to be considered on a plant-specific basis. Those actions

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were incorporated in the IPE program through Generic Letter 88-20, Supplement No. 1.

In Generic Letter 89-16, the staff requested that the licensees n

harden the suppression pool vent path under.the provisions of 10 CFR 50.59.

If a licensee chose not to make the modifications, it was requested to provide an estimate of the' plant-specific cost of the vent modifications. 'The staff planned to use these

Contact:

NOTE:

TO BE MADE PUBLICLY AVAILAELE Mohan C. Thadani, NRR 79 10 WORKING DAYS FROM THE i

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estimated costs to evaluate the cost effectiveness'of the--

requested modifications, and to determine if a basis existed for 1

the staff to issue Orders pursuant to the backfit rule, requiring that the modifications be made.

As stated in SECY-90-023, the licensees for five plants declined to make the vent modifications. The licensees for the remaining _

plants (exceptthePilgrimplant)volunteeredtomake'thevent H

modifications-(Boston Edison Company had previously made the vent modifications at its Pilgrim facility and was not required to respond to Generic Letter 89-16). The responses to Generic Letter 89-16 were sumarized in a memorandum, dated November 29

-1989, from Thomas E. Murley to James M. Taylor (Enclosure 1).

L Discussion:

This paper sumarizes the recent actions taken by the staff

-relative to the CPI program, and the current status of implementation of the modifications. The responses to Generic Letter 89-16 indicated that all but four affected licensees (five plants) intended to voluntarily make the vent modifications.

. A preliminary examination of the licensees' responses indicated '

that the licensees may not have firmly comitted to modify the vents. To clear up the uncertainty in the licensees' responses to Generic Letter 89-16, the staff reouested clarifications from all licensees for BWRs with Mark I containments.. For those licensees that had indicated an intent to install modifications, the staff wrote to ascertain if the licensees' declared intentions constituted firm comitments.

If d licensee's modification the directive of the-Comissioni,1993 (the schedule. based on schedule extended beyond January the staff requested an explanation for the delay.

If the licensee.had declined making'theLmodifi-cations, the staff made it clear that it was conducting backfit analyses for its plant. Those. licensees not volunteering were also informed that the staff will orovide the licensees the results of the beckfit analyses to give them another opportunity to consider their decisions in licht of the'results of'the-staff's plant-specific backfit analyses.

If the~1icensees continued to decline to make the modifications, the staff would:

require backfits in accordance with-the-provisions of:10 CFR' 50.109.

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None of the licensees who were planning to modify vents objected

.to the staff's understanding that their declared intentions to-sJ.

make the vent modifications, in fact, constitute comitments to s

~NRC. The. licensees whose schedules extended beyond January 1993 by a.few months, provided valid' justifications for delays of a few months in their completion schedules. Niacara Mohawk Power Corporation hes not yet responded to the staff's letter dated January 22, 1990. The staff is presently requesting a-prompt response from Niagara Mohawk Corporation for its Nine Mile 1

Point Unit 1 1acility, I

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'The staff has completed the backfit analyses for the non-volunteer.

licensees' plents. The staff's analyses calculated the reduction in risk that-would accrue from operation of vents for sequences involving transients associated with the loss of decay heat removal capability- (TW sequences). 'The staff has also' calculated the benefits of vent operation associated with the averted costs of cleanup of the site surroundings and the costs of-replacement

_ power. However, the staff has not quantified the small. risk reductions associeted with the operation of the vent system for anticipated transients without scram (ATWS)'and station blackout sequences.

In addition, the staff has not quantified the l

benefits of-suppression pool scrubbing for accident sequences r.uulting in core melt, t

Based on the results of the backfit analyses, the Lstaff has-l concluc'cd that the backfits for the five plants are warranted.

i because-the calculated reduction in risks due to TW sequences clearly indicates that the inste11ation of the vents will provide significant added protectior of the public health and i

saic,ty. Additional increase in protection of the health and i

safety of the public would. also result from operation of the vent-systen for the O) prevention of severe-accident sequences i

otherthanTWsequences,and(2)mitigationofsevere-accident j

consequences by_ scrubbing of the fission products in,the suppression 1

pool when core melts occur. The averted costs to clean site 1

surroundings and to replace power will be an added incentive for-vent modifications.

During'the rext few weeks the staff plans to write.to the licensees of the five plants not currently-planning to modify the_ vents, summerizing the results of the cost-benefit portion 4

of the backfit analyses, and requesting that theilicensees j

reconsider their decisions in light of the results of the staff's plant-specific backfit analyses. The' licensees will be given 30 days to respond. After:the 30-day period, the staff-will issue Orders requiring backfits for those plants whose

' licensees failed to respend or declined to revise their decisions.

The staff's target schedule for achieving the installation of

'l hardened vent capability on all Mark 1-facilities _was set for January 1993. The present status' indicates that'at least 13 of the 18 plants that will be modified on a voluntary basis will be-E modified by January 1993. An additional three plants will be modified by March 1993. Peach Bottom Unit 3 is presently i

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scheduled for modification during the Fell 1993 outage, and Nine p

Milt Point Unit 1 is scheduled for modification by March 1994 r

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The completion schedule of Farch 1993 for the three plants is l

E acceptable ~to the staff because the licensees have indicated i,

that-their outage schedules have constrained their completion j

schedule and because these plants will not be operating during i

g most-of the period between January and March 1993.- The staff l

will continue to negotiate further improvements in the completion schedules for Peach Bottom Unit 3 and Nine Mile Point Unit 1.

For the five plants for which the staff has performed backfit-j analyses.the modificetions will be required to be completed by l

p-January 1993.

As a separate action, the staff has prepared a Draft Generic:

EnvironmentalAssessment(DGEA)fortheinstallationandoperation:

of hardened vents (Enclosure 2). The DGEA will be published in the Federal F.egister for public coments.. The public will have-i a 60-day coment period. A11= coments received in a timely j

fashion will be addressed in the Generic Environmental Assessment -

before publication in the Federal Register. The DGEA has been L

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reviewed by the Offices of Fuclear Reactor Regulation, Nuclear Reguletory Research, and General Counsel.

The-DGEA is expected

-l to be published by r.id-Jui)e 1990.

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Enclosures:

I 1.

Memorandum from Thomat E. Murley-to James M. Tcylor dated tiovember 29,-1989 1

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Draft Generic-Enviror. rental Assessment

and Finding of No Significant Irpect j>

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DISTRIBUTION

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NOV 2 91989 l

I f.EMORANDUM F0r.: Jares M. Taylor, Acting Executive t

Director for Operations FROM:

Thones E. Murley, tirector

-b Oifice of Nuclear Reactor Regulation

SUBJECT:

MARK I CONTAlhMENT' PERFORMANCE IMPROVEMENT PROGRAM I

is:a follow-up to the Commission's directives related to the. Mark 1 Contair.nent Improver.ent Prograic,, the staf f issued Generic Letter 00-16 to all bcloers a

of licenses for reactors with-Mark I ocntainments.

In this generic letter, the staff strcncly encouraged the licensees to taate modifications to wetwell vents to herden the ver,t paths.

The staff requested that the modifications be made voluntarily under the provisiors cf 10 CFR 50.59.

If ar1y licensees elected.

not to make. the modifications voluntarily, they were recuested to prcvide their estimetes of wetwell vent modification costs, which could be used in the-staff's= backfit-anblyses for the plants of the non-volunteer licensees.

We heve now received the. licensees' responses to Generic Letter 89-16. Out of l

17 licensees (24 er its),12'(IE units) have iridicated that they will moke the j

- requisite modifications voluntarily.

Boston Ecison' Company (one unit) has 1

. The ren:aining four' already installed an acceptable wetwell modification.

licensees (five urits) have= provided.the estimates of their costs. and these estimates will be censidered ir the staff's-backfit analyses for their plants..

'The details of the licensees' responses are summarized in the enclosed memorandum.

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A resiew of the 'ivf enentaticr, schedules provideo by the volunteer licenstes l

indicates that several licensees will exceed the 3-year timeframe in which the Connissior, expected the hardened wetwell _ vent installttions to be completed.

The staff plans to contact these licer. sees and expicre the possibility ofi j

earlier l implementations so that all modificaticns will be completed by January 1

'1993.

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Enclosure:

As stated cc w/ enclosure:

All' Regional Adminuttaterc 1

Contact-M. Thadani, NRR 49-21427 A

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November 9 1989

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MEMORANDUN FOR:. Thomas E. Murlev, Director Office'of' Nuclear Reactor Regulation l

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James G. Partlow Associate Director #or Proiects i

o gulation Office of Nuclear Reactor e

SUBJECT:

LICENSEES' RESDONSES TO GENERIC L'iTTED 89-16 REL ATED _TO INSTALLATION OF HARDENED WETWELL_ VENT 1

We have received responses to Generic, Letter 89-16 from all Mark I licensees except Boston Edison Company.which has alret.dv installed an acceptable hardened wetwell' vent.

Out of the sixteen responses received, twelve have indicated that they will make voluntary modifications to harden the wetwell vent. They will work with the BWR owner's group to develop design criteria which will be submitted to the.NRC staff sometime in April, 1990. sumarizes the schedules of imolementation for the voluntary installation'of hardened vents.

l The remaining four licensees.have previded their' estimates of the modification costs and have indicated that installation of the hardened wetwell-vent is not expected to' be cost effective for their facilities.

The facilities for which cost estimatec have been received are (1) Fitzpatrick,1(2) Millstone, Unit 1, (31 Ovster Creek, and (4) Dresden, Units 2 and 3.

These estimates are provided along with the staff's estimates of eenefits in Enclosure 2.

Tb'e staff will perform the backfit analyses for the four.non-volunteer facilities and prepare backfit orders if the analyses support it.

The responses from the sixteen licensees are included as Enclosure ?.

M&

Jaqes G. Partlow Associate Director for Projects Office of Nuclear Reactor Regulation

. Enclosures As stated cC*

'F. Miraglia A. Thadani C. McCracken' J rudrick C. Li All Regional' Administrators cc w/o Enclosure a:

G. Gears.

M. Boyle-D. LaBarge R. Martin

P. O'Connor E. Tourigny T. Ross L. Crocker-

.i J. Stang.

A. Dromerick W. Long NRR Division Directors C. Shiraki D. Mcdonald G. Suh M. Fairtile All NDR ads All NRR PCs B. Siegel R. Hall i

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x ENCLOSURE 1 4_

SUMMARY

- 0F SCHEDULES FOR YOLUNTARY HARD VENT MODIFICATIONS AY MADK I PLANTS Plant Name' Modification Completion Schedule Browns Ferry 1 Prior to restart 9rowns Ferry 2 18 months from restart

- Browns Ferry 3 Prior to restart Brunswick 1 January 1993 Brunswick 2 January 1993 Cooper Sprina of 1993 Duane Arnold December 1992

~ Fermi 2 End of 3rd refueling outage (Fall of 1992)

Hatch 1 Spring of 1993 Hatch 2-Fall o' 1992

' Hope Creek January 1993 Monticello.

March 1993 Nine Mile ~ Point 1 March 1994 Peach Bottom ?

Fall of 1992

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Peach Bottom 3 Fall of 1993 Cuad Cities'l January 1993 Cuad Cities 2 January 1993 Vermont Yankee End of 1992 refueling outage 7

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1 ENCLOSURE 2 q

LICENSEES' ESTIMATES OF COSTS OF HARD VENT INSTALLATION AND 3TAFF ESTIMATES OF BENEFITS'

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.Fitzpatrick has two RHP loops. two pumps per RHR loop and one heat exchanger per RHR loop.

With the installa frequency reduction of 4.5 x 10'gion of the hardened wetwell vent, a core melt per vear. was estimated by the staff for the

' Fitzpatrick type of plant.

The licensee has estimated the modification cost for hardening the vent to be $680,000; and if the design were made ac independent there will be ar additional cost.of $70.000 Millstone. Unit 1 has one loop o isolation condenser.

The core melt frequency reduction of 1.4 x 10'{ per year was estimated by the staff for the Millstone, Unit I type.of plant. The licensee has estimated the cost of harriening the vent +o be $1.1 million; and if the desion were made ac

-independent there will be an additional cost of $0.7 million The Oyster. Creek facility has two loo The core melt frecuency reduction of 1.4 x 10'gs of isolation condensers.

per year was estineted by the staff for the Oyster Creek tvec of plant.

The licensee-has estimated the cost of hardening the vent to be $2.1 million; and if the design were made ac independent there will be ar additional cost of'$0.8 million.

Dresden, Units 2 and 3 each have one isolation condenser loop. The core melt

'reQuency roduction of 1.4 x 10'g per year was estimated by the staff for the Dresden tvoc of units. The licensee has estimated the cost of hardening the-vent to'bc (1.0 million per uniti end if the design were made ac independent there will be an additional cost of $0.5 million oer unit, j

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ki TENNESSEE VALLEY AUTHom Ty CHATTANOOGA. TENNtssts M40t SN 1578 Lookout Place OCT 3 01999 U.S.~ Nuclear Reg;1 ster / Cormission 1

ATTN: Cocerent Centrol Oesk i

Wast.ingt:n, D.C.

I')555 Gentle en:

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-In the Matter of

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Docket Nos. 50-259 Tennessee Valley Authority

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50-260 I

50-296 I

BROWNS FERRY N' CLEAR PLANT (BFN) - RESPONSE TO GENERI J

"INSTALLATICN OF HARDENED WETWELL VENT" 89-l6-of Boiling hater Reactor (EhR)On Secte-ter 1,1959, the NRC issued Generic plants with Ma'rk, I containmsats of tne NRC Performance Imorevement Pr: gram. program for ciscosition of tne issue 1

This letter encourage: utilities to-voluntarily, install a har:ene: vent uncer 10CFR50.59.

Our. review ' f SECY 89-017 indicates that the primary benefit of the hardened I o

vent is the' reduction of ' risk associated witn the loss of decay heat removal secuence.

Tnerefore, it is the intention of TVA to install'a hardened vent to reduce tne rist.

cf the loss cf decay heat removal sequence at BFN, be w:rking witn tne BR 0aners' Group to develop generic design criteria for -

TVA.will tre narcened' vent.

available for NRC reslewIt'is anticipated tnat the generic design criteria-will te by April 30, 1990, Specific design details will be Oeve10:e: 'as:TVA c:toletes the a::ropriate portions of the Procabilistic. Risk Assessment for SFN tne vent and existing plant cesign.and studies the possibility of systems interac It is TVA's intent to install tne

. rar:ene: vent on Uni.t 2' curing the first scheduled refueling outage after,

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This refseling outage is currently scheduled approvimately 18 mentns restart.

after resta-t.

Such an accelerate: schedule will require ex ecited approval 1

of the BWR Caners' Group generic cesign criteria and optimum materials avalla:Hity.

If cifficulties witn cesign or materials develoo TVA~will cis:Jss the proble*s and senedule for tne modification with'tne staff, The 4

nar:ene: vent.will be instal!ed before the restart of Units 1 and 3.

It is the belief of TVA that the risk of the loss of decay heat removal L

se:.en:e is alreacy low.

Tne installation of the hardened vent will further re:L:e risk of this seQLence ard make a significant contribution to closing tne M;5 s Mark 'I Containment Performance Improvement Program,

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L Tr.e c:M tre-t c.etained in this letter is listed in the enclosure, l

An Eowat Opportunity EmployF e e sem +^ em***

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.i Please. refer any Questions concerning this submittal to Patrick P.-Carter, 8FN Si te 41 censing, (205) 729-3570.

Very truly yours, TENNESSEE VALLEY AUTHORITY Jhw.

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Mana#, Nuclea icensing and Regulatory Affairs'-

~ Enclosure..

cc (Enclosure):

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lMs=. S. C. ' Black,- As sis tant Direc tor /

for Projects TVA Projects Olvision

.U.S. Nuclear, Regulatory Come.ission OneyWhite Flint, N0rtn 11555 Reckville Pike Rockvi11e. Maryland-20852 Mr,: B..A. Wilson, Assistant Ofrector for'.Inscection Programs

TVA Projects Olvisico

-U.S Nuclear Regulatory Ccer.15ston Region 11

-101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 NP.0 siident= Ins:ector Src.ns Ferry N. clear Plant.

Route 12, Bo c637 Athens,1Alacama 35609-2000

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Comttments Contained in~ Letter-i o;

Hardened Wetwell= ver,t (Generic Letter:(GL) 8916i

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.l The hardened vent will be installed in 8FN Unit 2 during.the next refueling cutage'and before restart for Units 1 and 3.

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- OCT 2 71989 a.4 *escu '

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SERIAL: NLS.89 291 ww o..wea tniied 5:

4:es Nuclear Re5ulatorv Commission A!!ENT!CN: Docun,ent Control Desk Vashing:en, DC 20555 BRk'NSV! K STEAM ELECTRIC PLAST, t' NIT NOS

. C KET NCS, 50 325 6 50 324/ LICENSE NOS, DPR 71 6 DPR 62

, 1 AND-2 RISPONSE'To CESERIC LETTER 89 16 lNSTAL!.AT MN'0F A HARDENED VETWELL VENT Cen:.emeh-

n 5 i

_ 3.'S,e :e.be:J1. 1969, n p. ants;with/ Mark L con:ainmen:s of the NRCthe NRC issued G

3ne issuesf re rm the owners 1

i,.h{s.e: er en.ated :o the Mark I Containnen program for disposi: ion cf ceuraged utilities o voluntarily install a hardenedPerformance

....-.)..n vent under

'L: review cf SECY 89 Ol? indicates tha:

ven:

is.:ne redue: ion of risk asso i the primary benefi: of-:he~ hardened

[e :va.:

seq c a:ed with the TV -(loss of decav hea:

.02;&ny -r5yence.

Therefore, i:

is the inten ton of Carolina Powir'6 L;z; :

to install a hardened ven 1

sequence the Brunswick S:ea wi:n :ne a:

to reduce the risk of the TV

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Elec:ric-Plan:

SVR Owners' Croup :o :evelop generic design crite iCP&L'will b (BSEP),

d ven:;

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p.e te.s s:n ril 30. pe ted tha:

the design criteria vill be available for NRCr a:for th review by Ap

1990, Specific' design details will be developed as CPLL e appropriate portions of :he IPE for BSEP and studi

ssib!'.!3y 'of systems

interacti

>M f.gn" [d.ne nardened vent will be -ins:alled -in BSEP by either Ja es':he f-:

'c r 1 vni:never :pri[.5 the second refuelin5 ou: age from the date of :his s.a:er.

is :Me f:elief of CP&L :ha:

insta.,.'a:icn Of :he hardened ven::he risk of :he TV sequence f.s already lew The 3n:,tase:asjgnift:antcontribu:!onto-:losingtheNRC'will fur: hor reduce risk of th ter::rian:t i

.:provement Program.

s Mark 1 Con:41 n,me r.: -

if; s

y:y,speu'd have any questions. please contact Mr Stephen D

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Yours.very truly, l-O N

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R..A.

Watson

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Mr. S. O. Ebne:er p

• r. 2 M 'Ruland Mr L, d. Thornburg Mr. E. C. Tourigny o

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f Subte::: Resper.se to Generi: Letter _89-16 d

nsta.',ation ef a Hardenec Wetwell Vett

per S;;'. ear Sta:lon

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.ny, ne Nuclear.regu..atory Comis61on (NRC) issued Gener;;

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Tne purpcse cf GL 99-;6 was to infert the owners of Sciling-Water Fea:::: ( Shi) p ' a r. t s.;;h Man ! ::ntalments of the NE; program for disp s;;;;r :f the ;ssues related to inc Mar <.: Contair. ment Performance

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mpr: cerer. Fr:gra. ih;s gener; let ter en:curaged utilities to voluntarily-

r.sta...a narder.ed vent u., der

ne pr: visions of 10 OTR 50,59.

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A-rev.e. :f SE rl 59-01: ty the EW7

.;;ities indicates tha:

ne preary l

renef:

f tre tardened ven, is :ne rec..::en of risk associated witn t,e ;;ss-f de:a

'reat rex va.'

se:uence ( 74 )

a; bciling water react:rs witn Marx g

";nvenir.

Tr.eref:re. :ne Se:rasxa P::.ic Power Oistrie: (:ne :::s:::: >

.r. ends :: ;rstal. a nardened ver.: a: the Copper -Nu:; ear Station :CNS). !ne J

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.;.. :+.Orking.;;n :ne SWR Owner's Group :: deve. p generi: des; will also re.'y upon :ne resu;gn j

y :riter:a f:

.. + ;.stalla:::n.

Tae Ois*ri::

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! ! ne ::.d ;v i du a '. Plant Exa.inatier ;!PE) study fer OSS, to identify sys;e-

-N.: era::::r. ef f e::s :;etween :ne ha.rdened vent and ne existing plant des:en.

.s ar*.;;;;ated.ha

ne gener:0 BWR Cwner's ' Group design criter.a wi.1 be

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avs..ar e f:r SR; review cy Apr:1 3*,

1990.

The Oistrict's :ndividual P' ant Exa.n.a:::. s:heculed to be su:mitted in October 199;, will ine'ude :ne

' effe::s

f installing a hardened vent en the Cooper Nuclear Station.

i ineref:re, since there is no refueling outage seneduled in 1992, the Distrie:

nter.ds to install. a. hardened wetwel.' vent at the Cooper Nuclear Statien by

' l s.artup f ree :ne 1993. refueling outage, curren:'y scheduled for the spring cf

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- ;s :ne ee';ef cf tr.e Nebraska Pubil: Power 01strie: rhat the ::sk of.tne

.':ss :f de:ay neat remova; sequen:e ( "'W ) is aiready low.

The installation of t r.e

.ar:e r.e : vent may further reduce tne risk of this sequence and -11.

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.Or:riru.+ :: :.0 sir, :ne NRC's Mark : C ntainment Performance Improve.en; i r:.g r a.

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= s hv. '. d y:u r. ave ar.y aestions -concerning this response, please contact' this '

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Stt:erely; i

G. A..rev:rs tiv;si:r Mar.ager

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lowa Electric Light and Power Company October 31, 1989 NG-89-2886 Dr.. Thomas.E. Murley, Director

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Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Attn: Document Control Desk 1

Mail Station P!-137 Wasnirgton,- DC'20555

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,Sueject: -Duane Arnold Ene gy Center Occret No: 50-331 Oe License No:

DFR-49 i

Resoonse to Generic Letter 89-16, " Installation of a Har:ened Wetwell Vent"- -

Refe er e:

Generic Letter 99-16, " Installation of a Hardened Wetwell Vent," cated Se:temoer 1, 1989--

File:

A-10lb,_A-106a, A-107c, T-23a ea* :. v

'ey:

u i

~Ge*ef:.ette- (GL) 59-16 re:cested that ea:h if:ensee with a Ma k I certairmeat

te :e y:ur sta wit. i ts clans for instaat ca~ of a har:ened wetweil i

vent ae: ase' ele:te: to cec:ee: with the cesi;* aa installation of a narceae: vert

~

ir:e' t e : : visions of 10 CFR 50.59.

ae. aie a :: :e:tual cesien of a hardeaec wetwell ve~nt f:r the Duane Ar9c10 E er';> :e te- (DAEC) tnat we celieve accresses -ine Commission's concerns for eeu:t': Of tre ist associatec with a TW (loss of cecay. heat removal) accident.

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se:Uea:e.

A trief description of this coacectual design is.provided in Atta:- eat 1.

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fj Gene *al OfGce

• P O BM 33 8

• Lew Macods lows 32406

• 319'398 4411

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.Dr. Thomas E. Murley i

. October 31, 1989-7,. L NG-89-2886 Page 2' Iowa-- Electric new anticipates completing the hardened vent insta11ation'by-

)

. December 31, 1992.

We will keep you informed of our plans, milestones and

'seneaules through the semi-annual updates to-our Integrated-Plan,

o-t Very truly yours; f

Daniel L, Mineck e

Manager, Nuclear Division DLM/BHJ/:Jv+

I Attachment i

c::

B. Jennsen L, Liu

'L.-Root.

R. M:Gaugby

'J;'R. Hall (NRC-NRR) m A. Bert' Davis (Regien III)

NRC Resicent Office 2Com-itment Contrei 8890346 t

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Attachment I to:

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NG 89-2886 Page J of 1

o CONCEPTUAL DESIGN OF THE DAEC HARDENED WETWELL VENT-Our conceptual design for the hardened wetwell vent will connect the existing piping in tne. reactor building for the wetwell purge exhaust to existing piping in. the turbine building for the discharge from the steam packing exhauster (see Figure 1).

The discharge piping for the, steam packing exhauster-leads to-the plant offgas stack, providing for an elevated release point.

s:'

p

>The new piping will be eight inches in diameter and will' connect to the wetwell

. purge exhaust piping between the existing inboard-and outboard torus purge and vent' isolation valves. ~ An additional air-operated outboard isolation valve and

a. rupture disk will be installed in the new piping.

The design will allow'the vent to operate during station blackout event, 1.3, it will be AC independent

. commensurate with the duration specified by DAEC's compliance with the station i

blackout rule.

The new isolation valve will fail-safe on loss of air, but will have a compressed air accumulator to support valve operation following a loss-o of~ instrument anc service air.

Controls for the existing inboard ~ isolation-valve will be modifiec to allow it to be operated indeceacent f AC ::wer. Valve control anc' position incication will be provided in~the conteel reom.

i

'The new ciping from the ta on the wetwell purge exhaust piping through the new isolation valve will be ASME Section III, class 2 and.will be seismically succerted. The remaincer of the piping will be non-safety grade 150 psig carbon steel piping.

7 o accitional snielding crevisions are ensisioned.

Also, existing eauicment N

for mea'toring raciation in the drywell, wetwell, anc of fgas: stack will provide 7

'c.apaed 'ity 'er this crocosed installation, s

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Octcher 20, 1989 p

NE-89-0216 t.

U.. S. ibelear Regulatory Camission Attns Document Control Desk Washington, D. C. ~ 20555 i

References:

1)

Fermi 2-NK Docket No. 50-341

^

NK License No. NPF-43 j

2)

N)C Generic Letter-89-16, 4

R

" Install.ition of a Hardened

. Wetwell Vent", dats$.

Septa =ber 1, 1989

Subject:

Assoonse to Generie htter 89-16 t

his letter is to provide Detroit Edison's-response to NK Generic Intter 89-16 (Reference 2) which was received on September 7,1989.

The Generic Mtter encourages all. plants with Mark 'I Containments to voluntarily install.a hardened wetwell vent under 10CFR50.59, or s

provide a plant-specific cost estirate for inplementation so that the NK can perform a plant-specific backfit analysis.

Detroit Edison has decided to proceed with plant modifications to

. install a hardened wetwell vent under the provisions of 10CFRSO.59.

Detroit Edison is working with the BWR Owners' Group on the proposed hardened vent.

u he esticated schedule for. coupletion of these-modifications is the erd of our third refueling outage; this sche $ule will be reviewed following deterr.ination of:the finalized scope.

i If you have any questions, please contet Mr. Girija Shukla at (313) 586-4270.

Sincerely, l

.4

[ylhi.

I da ec: - A. B. Davis R. C. Knop.

al' W. G.~ Rogers J. F. Stang L

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HL-782 0334V October 2t.,

1989 i

U.S. Nuclear Regulatory Commission d

ATTN:

Document Control Desk M

Washingten, D.C.

20555 PLANT HATCH - UNITS 1, 2 NRC 00CKETS 50-321, 50-366 OPERATING LICENSES OPR-57 NPF-5 l

GENLRIC LETTER 89-16 J!

HARDENED VENT Gentlemen:

1 On Se: tem:er -1,1989, the NRC staff issued Generic Letter 89 ::

inform :ne owners of BWR plants with-Mark I containments of the NET

r: gram f:r discosition of the issues related to the Mark I Containme :

i

e 'or.an:e I ;rovement Pr: gram.

This letter encouraged licensees :o j

1 voluntarily install.a hardened-vent uncer 10 CFR 50.59.

Our review of SECY 89-017 indicates that the-primary benefit of the

• ar:ene:

~ ':ss of de:ay heat ' removal) vent is the' reduction of risk associated with the TW-(lo i

Dewer Com:any to install a hardened It' is the' intention of Georgia 1

sequence.

se:.en:e at Plant E. I. Hat:h, Units l' and 2. vent ~ to -raduce the risk-of the TW Georgia Power Company will.

=:e a:-d eg wi th the BWR ~0wners' Greu: to develop generic. general-at:-

5:i: 't: :esign criteria.

It is anticipated that the design criteria will te available for NRC review by April 30, 1990.

Specific ' design

stai!s

. il1 ce available as Georgia Power Company completes' tne w

a::r::riate portions of 'the IPE for Plant Hatch and studies tre

-::ssitility of systems interaction effects between the vent and tre--

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existing plant cesign.

The hardened vent will be installed on Unit 1 te':re startup from the end of cycle 14 refueling outage and.on Unit 2 te': e start.c from tne end of cycle 10 refueling outage.

It 's t'.e Oeiief of Georgia Power Com:any that the risk of the *W se:'en:e is alrea:y low.

The installation of th6 hardened vent will

'artner ee: :e risk of inis secuence and-make a significant contribution

cl0 sing tne NRC Hark I Con:alnment Performance Improvement Program.

1

I GeorgiaPouer A

-4 U.S. Nuclear Regulatory CommissionL October Oa. 1989' Page Two.

l

'If you have any questions in this regard, please contact this office, Sincereiy,

.M)./b,' h C

W. G. Hairston, III JKB/eb' c:

Geercia Dewar Cemraav Mr. H. C, hix, General Manager -' Nuclear Plant Mr. J. D. Heict, Manager Engineeri.ng and. Licensing - Hatch GO-NORMS U.S. Nuclear Degulaterv Cemmission. Washincton. D.C.

.Hr. L. P CrocKer, Licensing Project Manager - Haten i

U.S. Ne:1en-Reculaterv-Com-ission. Recion II Mr. S. D. Ebneter, Regional Administrator Mr. J. E. Menning, Senior Resident Inspector Hatch t

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.s m o NUCLEAR FUEL CYCLES PLAN OF. RECORD i

1989 1990 1991 1992 1993 3

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Based upon NTC acceptance.of the. generic criteria 'within 6 months W

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=of. submittal by-the BWR Owners' Group, any upgrades to our 1

. existing-six inch hardened. vent will be completed prior to 3

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' restart-following the!second refueling outage from the date of.

i 1h.

this letter or January 1, 1993,'Which ever is later.

It is;the belief of PSE&G that the risk of the TW sequence is Jalready

low.

Upgrading the existing hardened ~ vent will further reduce the risk of 'this sequence and make a significant contribution to closing the NRC's Mark I Containment Performance F

Improvement Program.

Sincerely,

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,w, 4

se C-Mr.

C. Y. Shiraki Licensing Project Manager Mr.

D.

K. Allsopp.

4 Resident Inspector Mr.

W. T.. Russell, Administrator Region I Mr. Kent Tesch, Chief Bureau of Nuclear, Engineering New Jersey: Department of Environmental Protection-i a

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NLR-N89220 1

-STATE OF NEW JERSEY.

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COUNTY OF-SALEM

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do La Bruna, ~ being duly sworn according to law deposes and saysi-i I am-Vice President - Nuclear operations of Public Service j

Electric and Gas company, and as such, I find the matters set 1

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forth in our letter dated october 30, 1989

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concerning the Hope ~ Creek Generating Station, are true to the best of my L.-

knowledge, information and belief.

W y'

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e Subscribed and Sworn to.b o e me i

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this day of 1989 f

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$dtary Pubid of New Jersey l W.i LARAINE

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l 8kr9eam Stumn Peeer Conipany staI*eemet Men Emmeesens. Enneesta ste011627 Toooaone totri tateco October 21, 1969 Ceneric Letter 89 016 Director of Nuclear Reactor Regulation t' S Suclear Regulatory Commission A::n: Document Control Desk Vashingten. DC 20555 MONT: CELLO 14' CLEAR CENERATINC P! ANT Decke: No. 50 2(3 License No. DPR.22 Respense to SRC Ceneric Letter 89 16 k atallatten ef a Hardened Verve 11 Vent Cn Septe.ber 1,1969, the NRC issued Generic Letter 6916 to infor= the e. ers of Bei'.ing Vater Reactor (5*.7) plants t".th Mark I containments of Commiss :n decis: ens regarding issues related to the Mark 1 Centairaent Perforzante i

' prove::en: Frogra=.

That letter encouraged utilities, on their own initiative to install a hareened vent under the provisions of 10CTR 50,59 and reques:ec a response within !.5 days. This letter provides our response.

Cur reviev

',f SICY 89 017 (Mark 3 Centairment Performance Program), referen:et in Gent %: Letter 89 016, indicates that the primary benefit of the hsrdete:

is :..e reduction of risk associated with the TV (less of decay hea:

ven:

re::va'.? sequence.

It is our intention to install a hardened vent co redu:e the risk of the TV sequence. Ve vill work with the BVR Owners' Group.to develop generic design criteria.

It is our intention to develop a plan:

spe:ift: design, based upon the generic criteria and any appropriate inputs iden: fied during the Individual Plant Evaluation (IPE), which vill consis: ef a d,edicate,d vent routed to the top of the reactor building rather than specia'

)

adaptatien of existing systems to provide a route to the stack. The hardened.

ven: v '. '. b e installed prior to restart following the second refueling ou: age frc: the ca:e of this letter, currently scheduled grch 1993.

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Northem Statee Power Company Directer of bTJL October 31 Tage 2

'Je vill nexify you if we determine a need to change our intentions or completion schedule as the generic criteria, IPE inputs or plant specific i

criteria are being developed.

Please contact us if you have any questions related to this issue.

4 177M W.4J Then.as F. Parker F.anager

)

.'?uclear Support Services c: Regional A1.inistrater. :::. 57C b7.R Proj ec t' F.anage r, b'RC Resit.ent inspector, 57.0 j

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October 30. 1989 NMP11. 0447 J.S. Nuc'est Ee;s'at:ry Commission

]

atin: Doc. rent cntesl :est aasniagton. 0.:. 20555 t

Ee: Nine Mlle Point Unit 1 i

Doctet No. 50-220 OPR-63 i

l entlemen*

i

• • is attee 's in re:1 y to Generic ' etter 89-16 date: Sectemcer 1,1989 regar:ing esta11ation Of a Har enec Wetwell Vent, feur letter statte that t*e Commissten as :irecte: tre Sta'f to accreve the Installat'On Of a nar:eaec.ent.":e

e :r: visions of 10 CFR 50.59 for licensees.no elect to I

• ac:recrate

• >s O'n't M:rovetent.

Your letter concluces t94t a reliable

)

• arceate met.eli < eat all:ws for c:nsi:erati:n of coercinate: acci:ent i

'aan;teemt strategies :y :rovicing :est;n cacability consistent alth safety

jectives.

l Niag4*a MO*4nk, wi'1 ensure provisions esist for Wetwell venting at Nine d 'e 80 int unit No. I by evaluating entsting systems for aceouacy or cerformin; ecessary x ci cations.

The criteria nich will ce usec to cerform ine eva' atien i

nst:er licensing bases anc [mergency Ocerating 8rece:ure impietentation recuirements, ModiflCations will be.installe0 uncer the j

rovisiens :f ) CFR 50.59 uttilzing portions of eilsting systems to tnt-

• estest eitent cessible. Our estimates scnecule for installation of Mcificaticns snoul0 tney De necessary, is the sec;nc refueling outage after

gr carrect : stage, anicn is tentatively 5:neculed for Maren 1994 Very truly jours, N:AGaRA MOHAW WER 00RPORATICN f

3 CT0.T9ery Vice Pres 10ent Nuclear Engineering anc Licensing

• S / *10 791aC se:

Regiend' administrator, Region I Mr. 4. A. CADra, Director f.'

Mr. R. E. Martin, D'oject Managed" Mr, W.

A. Coct, Resi:ent Ins:ec tor Re::r:s Management 1

)

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GL 49.;6.

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PHILADELPHIA ELECTRIC COMPANY c.

l NUCt. EAR CiROUP HEACQUARTER$

985 48 CHESTEP:9R.)OK BLVD, WAYNE PA 19c8M691 g mu, taisi see sese suscweeve viss eewes=, ave 6 san I

October 30, 1989 Docket Nos. 50-277 50 278 License Nos. OPR-44 DPR 56 i

U.S. Nuclear tea

-u atory Comission ttn: DocumentdontectDest Washin9 ton. DC 20$st h

$UBJECT: Peach Bottor itomic Pov'er Jtt.t on Unitt 2 and 3 Generic te:ter 89.'6, ':nstt'latiori of a Harcened Wetwell Vent" Gentlemen:

i NRC Generic Letter catec Septetur 1. ;999, recu(CL) 89-M "Installatic i of a Hercenec Wetwell Vent,'

etsconte which prov cts ')otif'catiot, cf our pla*s 'o accressing th'rtc Philactlpnit. Ele:t'ic Company vent issuti.

1 GL 89 16 t ractic lictrsees with Mhrt e narcened wetwell

-l with plant moeificatisnc ane trovict. an estitatid icnecule in the response; p plants to volunterfly i

ot herwise provice ',ott estimates fer inclementition of a nordente vent by pice replacemen,t for NR; staff ws4 in performing cla't.;pecific backfit a i

Our responte. :rovided in the attacuart' provides notification that we

~

proceec with plant dMi#ications to im: rove tse cut rett venting capabilities at P Bottom Atomic Power Station.

The esta))isamet od critcria anc'schecale for c

assessing and implementing potential m.)cific.ition; it, described in the respons 1

If you have any questions, or require 4.cd*tional information, please us.-

Very truly yours,

,A

-1 Attachment i

W. T. Russell, Adm9nistrater, Pegicn I. USNR; ce; T. P. Johnson, USHRC Sen'or Resicert Inspects

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.e y p1 ATTACHMNT PEACH 90T*0M ATorlC POW (P, STATION, UNIT $ 2 AND 3 RESFONSE 10 GENEP.!C LETTER 8916 i

"!NSTALLATION (F A MAPDENED *ETWELL VENT" l

are %e recently M11thec GEG l'50, entitled $tvtre Accise for t'ive U.S. Mucitaa An Assessment Power Flants.

NOREG-1153 provides a state-of the-art understancing of severt accicent r*sk and also povides NASH-1400.

is extremely low.

Cha ges in plant ::nfigurattoi and procedures, the evolution of Procatilistic Risk Assessment (8AA severe accident pnenoment have all)csstributte t)atthocolog,,, and an increaten uncer core casage frecuerney ;C0F) from trat in WAH-l400 to that in NUAEG 1150.a fa the most cominant scenario from dA$H 14DO, :Se lots of decay heat removal (T In fact, decreased tnree orcers of magnitade due to t morn realistic assessment of centa venting using existing equipent an: tu:cest'u Nottion following venting.

Peach Bottom hat 'melenentec Revision 3 of ne BWCG Cmergency Operatin (ECP) Guidelines *nita incorp tu;e ttm ust ud Detallen emergency prxttcures eex'st fc* each of the nine icentif c ntainnent.

on the emironnent, personrel, enc.tcuipment. vent paths ano are tricri Tra hard-pipec vent paths from the torus are prioritired first since tier prov*ct tre bett choice for satisfying the criterit of a scrubbec elease w'tn l'ttle inact on personnel anc equipment.

hard-pired 6 inen Integ*atte lett Ritt Test (11AT) flow path is the principal vent The oath catable of handlirg depressurization fitw *stes tssociatec with decay heat.

This particular flow pain crigirate; f rom tre witteel' airspace and cischarges O tre reactor Duileing.

I The emergeacy procedures regarcirg serting were used as the basis in huREG-11 4

determining the er:baci'ity of f ail'ng t,o successf ally implement the recuirements fo venting. Given nt time of 1 in 100 (.01).as use,d to repretent operatorpr:ceduret, are barewhrt available, a failur f 111ure,

in aceition, an extensive NRC revita Operating Procecuri Ins:ection (50- m (8)/88-200)of the ven:19g caca L

in August, 1988. The inspection j

team corcluded that PEC: was capabit of carrying >ut the provisions of the E0Ps

(

concerning primary containment venting using exis; ng equipment except under the special :encitions associated with ststion blackout.

Electric Company (PEto) wiPi proceed witi plant moeifications, NUREG-CR-5 25 Addencum ' and SEC'I 8g-017 inoicate the greatest risk reductionThe a

-petential frem installation of a narsenes vent 's tchievec in reducing even (vrtner the crocacility Of t5t postdated loss od i

little crecit for existing venting capneliity. decay neat *emoval scenario (TW) assuming clean steam vent) 45 the assessment basis whern ettermining the risk reductio potentia' of mocificttient at Peach 30tt:m.

L Qaner's Group to ceveio; generic design crittrit fer *he harcened vent.PE() will be anticipatec that the cesign criteria w'1' be avellable for NRC review by A(ril 30,

! is 1990.

Seecific cesign catails wi'1 se develcpec as PEco completes the appropriate p

pertions Of the incivicual Plant Examiration (IFE) for Peach Bottom anc studies the

]

cessibility of systens interacticn e#f tets betweer tSe vent and existing plant oesign.

Concuctec cwring the peach Bottom IPE ; recess.Evalutt'en of containmert v l-

- ~.

i' PICo Response to GL 89-16 l

Attaennent Page !

4 Caress anc redWCe the appropriate $svert sCcid a plent soecific etsis of assessing the most effe v

ng i

4 continute Ptto oesit'en of previcirg ano ennan:ctive modifications. This maintains l

health enc $4fety, In; the protection of the pwblic i

The modifications will be inclementes prior to restart following outage (Relesc 9) at etch unit.

These outages era currently projected to occur in l

the f all of 1992 #cr Urit 2 and fall :f 1993 fer Unit 3.

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Comm;nwealth Edleen

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2 *est Acams Soest CNcaso cros l

7 Ace' ens M>y to Pos omne Boa 7s7 i

/ CNca0s. Hwes 60600 0767 October 30, 1989 b

i U.S. Nuclear Regulatory Commission ATTN:

Document Control Desk Washington, D.C.

20555

Subject:

Quad Cities Station Units 1 and 2 Response to Generic Letter 89-16 NRC Docket Nos. 50-254/265

Reference:

Generic letter 89-16, Installation of a Hardened Netwell Vent, dated September 1, 1989.

Dear Str:

The referenced Generic Letter informed the Itcensees of Bolling Water Reactor (BWR) plants with Mark I containments of the NRC program for disposition of the issues related to the Mark I Containment Performance Improvement Program.

The Generic Letter encouraged Itcensees to voluntarily install 4. hardened vent under the provtsion of 10 CFR 50.59, and reovested that licensees provide notification of their plans for this issue.

This letter provides Commonwealth Edison's (Edison's) response to that request.

SECY 89-017 indicates that the primary benefit of the hardened vent is the reduction of risk associated with the TN (loss of decay heat removal) secuence.

To address this TW secuence issue, it is Edison's intent to provide a hardened vent at Quad Cities Station Units 1 and 2.

Edison is participating with the BWR Owners' Group (BWROG) in the development of generic design criteria for.the hardened vent.

It is anticipated that the SNROG will provide this-design criteria by April 30, 1990, for NRC review. Plant specific design features will be developed subsequent to formulation of the generic design-criterta in accordance with the 10 CFR 50.59 process to ensure system

)

interaction issues are addressed.

The hardened vent will be provided for Quad Cities Station Units'1 and 2 by January 1, 1993.

Edison will be performing an Individual Plant Examination (IPE) for Quad Cities Station to fulfill the reautrements of Generic Letter 8b-20 with an expected completion date of June 1993. Upon completion of this IPE, a review of its results may identify additional design and operational considerations specific to Quad Cittes Station with respect to the hardened vent.

0296k: 4

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U.S. NRC 2-October 30. 1989 I

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1 It is Edison's understanding that the provision of the hardened vent

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will make a significant contribution to closing the NRC's Mark 1 Containment Performance Improvement Program.

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Please direct an l

response to this office. y Questions that you may have concerning this j

Respectfully,

}I{ 'b M.H. Richter Generic Issues Administrator A.B. Davis - Regional Administrator. Region !!!

cc:

Resident Inspector - Quad Cities i

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VERMONT YANKEE NUCLEAR POWER CORPORATION l

g FerevAca: S tan'eDoro. VT cs3017002

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Et,31t,EE AP.3 0Fr ;g (p

m w. ru t-v e:. v, m : c i:5w October 30,1989 m

BvY 69-99 United States Nuclear Regu! story Commission Document Control Desk Washington, DC 20555

References:

a.

License No. DPR 28 (Docket No. 50 271) b.

Letter. VYNPC to USNRC, FVY 88 16, dated 3/1/88 Leiter. VYNPC to USNRC, BVY 89 83, dated 9/1/89 c.

d.

Letter USNRC to VYNPC, NVY 89191, dated 9/1/89 (Genenc Letter 8916)

Subject:

Installation of a Hardened Wetwell Vent

Dear Sir:

Letter 5916 Reference idRThe puraos: of this letter is to comply with the NRC This update indicated our plans regarding installation o r

protection capability (har&r :d wetwell vent) at Vermont Yankee, Reference (c) also pro conceptual description of tne planned vent, along with Vermont Yankee's commitment to overall plant safety.to suppon and follow ongoing efforts to resolve severe accide On September

.which transmitted the following instruccons:15,1989 Vennent Yankee receive i+

"The NRC staff requests that each licensee with a Mark I plant provide notification of its plans for addressing resolution of this issue. If the licensee elects to voluntarily i

proceed with plant modifications, it should be so noted, along with an estimated l-L schedule, and no funher information is necessary. Otherwise, the NRC staff requests that the above cost information be provided. In either event,it requests that each licensee respond within 45 days of receipt of this letter,"

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As stated in Reference (c), Vermont Yankee expects to establish specific design criteria that we can install enhanced containment overpressure proteccon capability by the end of t refueling outage.

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'4 implementation. If you have any questions or concem Very truly yours.

VERMOST YANKEE NUCLEAR POWER CORPORAT ka.o Leonard A. Tremblay, Jr.

Senior Licensing Engineer 1

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USNRC Region ! Administrator i

USNRC Resident inspector YYNPS USNRC Project Manager. VYNPS

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uthority-i a

Oct0ber 27,1969 J8N 89 70 '

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U. S. Nuclear Regu! story Commission Anention. Document Control Desk Mail Stop P1 137 Wasnington. D. C. 2055.5 S.:e;e:::

James A. At2 Patrick Nuclear Power Plant Docket No. 50 333 Generic Letter 8916 Instanation of a Harcened Wetwell Vent i

Ae'erences.

1.

NRC Generic Lener 8916 dated September 1,1989, James G. Parlow j

to NYPA rega'c.ng the same subject.

2.

NYPA letter, J. C. Brons to NRC, dated January 12,1988 l

(JPN 88 001/IPN 88 001) regarcing revisions to comme',ts on NUREG 1150.

3.

NYPA lener, J. C. Brons to NRC cated September 28,1967 (JPN 67 051/IPN 87 045) regarding comments on craft NUREG 1150.

Cea' Sir-T e Auth:rty has reviewed Generi Laner 8916 (Reference 1). This letter and its attachments 1

5at s'y t*e stats re: gest that ea:h Marn I licensee inform the NRC of its plans to install a

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• a':e e: :enta:nment vent. For the reasons outlined below and further detailed in Anachment 1.

i t e A,;tnergy will not volunteer to install a harcened vent at FitzPatrick at this time.

1 9st, the NRC statt has not justified why this issue should be given unique or special treatment

ataer. it should be resolvec in the same way other SECY 49017 issues are being resolvec. as -

Oa ; cf the IPE/PRA (Individual Plant Evaluation /Probabilistic Risk Assesament) currently in Or:;tess.

Se::nd the Authorrty's current analyses, together witn the unique circumstances and features

' the Et:Patt.:k piant. c0 not justify insta'!ation of a hardened wetwell vent for the TW secue.:e-ine Generic Letter inappropriate:y prescribes a generic mod:fication for a clearly plant.spe:if.:

se.ere a::icent issue.

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t TNrd, SRCY4C17 (upon which Generic Letter 8516 is baserf) contradicts both itself and other NRC sponsored studies on several technical points.

J Integrotion in IPE/PRA Progrom

)

1 The NRC staff has not adeoustely demonstrated wtg the herdened vent issue thould be l

resolved indepencent of IPE/PRA programs. The other ' potential Mark I improvements' mentioned in SECY SS 017 were subsumed into the IPE process No account for possible unique cesign oferences that may bear on the necessity and nature of specific safety improvements.'

l Stuo:es prepared for the NRC make it very clear that a containment vent is equally plant spe:if.:.

l NUREG.5225,'An Overview of BWR Mark 1 Containment venting Risk implications? Addencum 1 l

states:

• Applying the Peach Bottom results to other plants requires careful consideration. The sequence frequencies, local population dens.ty, evacuation plans, and plant unique features could have a -

sign:f: ant impact on the resutts.' (Executive Summary)

The ana'yses anc justJcat< ns referred to in the Generi: Letter and SECY 89 017 do not

ns.:er severa! reievant issues, a:temative approaches or near term improvements that s gnit:aa.t!y reduce the frecuency and consequences of severe accidents. Other modifications or a

proce:ura, changes with greater benef.ts and lower costs, should be considered before oe:id;ng en a hardened vent.

The Authority has significant experience developing and using PRAs. One of the first utitty.

s:ce,screc PRAs was completed a:most ten years ago for our Indian Point 3 Nuclear Power P:a-t as paa cf the Zion /Inc.an Point PRA stud.es. The AQorrty also has significant experience in tre ana ys:s Of severe a::icents having incepencently prepared and submitted extensive comments On NUREG 1150

• Reactor R:sk Reference Document,' (References 2 and 3).

The Autheitty's previous expenence indicates that PRAs are systematic, scientific tools for i:ent@ing : st effective enanges to improve safety at commercial nuclear power plants. We futy exce:t to ga:n insignts into those structures, systems and components that contnbute mest 1

s gmt: ant:y t: risx when the Fit 2Patnck IPE/PRA is completed. Any decision to install a M9eae:

vent bef:re an IPE/PRA has been completed is premature and only serves to undermm ge iPE process. While the Authority may eventually determine that some type of vent at Fit 2 Patrick is i

useful, su:n a choice should only be made in the context of the larger IPE/PRA analysis.

L Severe Accident Analyses The 45 days a' lotted by the Generic Latter for a response was insufficient to fully develop and eva'vate a harcened vent or the multituce of attomate approaches and mitigating features not consicerec by the sta*f. In the limited amount of time available, the Authonty has developed a p sition paper wnich summ:tizes some of these points, a copy of which ts included as. This caDer also identifies severalinconsistencies in the approach prescribed by tne j

Genen: Latter anc SEOY 89 017.

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MMed Vent Cest Eggmate Attachment llis the Authonty's preliminary cost estimate for installation o based on the 646Cription included in Genene latter 51C, 5404W00 only respond to the Genene Letter, the Authority reserves the right to reene the estimate.

contact Ms. S. M. Totn of my staM,Snould you or your staff have any s

Wry truly yours,

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John c. eron.

Executive Vice President Nuclear Generation ec:

U. S. Nuclear Regulatory Commission i

Reg:en i 475 Moncale Road King of Prussia, PA 19406 Ctfice of the Resident inspector U. S. Nuclear Regulatory Commission P.O. B0x 136 Lycoming, NY 13093 Mr. Dave t.aBarge

' Proje:t Diroctorate 11 Division of Reactor Projects. l/ll

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U. S. Nuclear Regulatory Commeon YaJ Steo 14 B2 Axkvil:e, MD 20555 J

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e Attachrnent 1 to JPNM 1

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Response to USNRC Generic Letter No. SS16

• Installation of a Hardened Wetwell Vent

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l New York Power Authorrty t

James A. FitzPatnck Nuclear Power Plant

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l Docket No. D333 L

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Table of Centen3 t

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SUMMARY

11.

BASES REYlfW OF SECY49417 A. Cost / Benefit Analyses 1.

Basic Equation 2.

Present Worth Analyoas t

3.

TW Segance Person Rom Agures 4.

Population Effects

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5.

Net Risk Ana!yses l

6.

Summa y B. Alternatives to Hatcened Vents 1.

CNeNiew 2.

Artt Steps 3.

Altemative Means 4,

Ftt2Patnck SpecM Stud <es 5.

Summary tv.

REFERENCES V.

TABLES AND FIGURES l

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SUMMARY

D M Will not volunteer to install a hardened verg a pipetnck a this time posrtion as based on severalpoints.

First, the NRC statt has not justified why this issue should be given unHlue o

- treatment. Sather, ft should be resolved in the same Fit 2Patt:ck show6c not be made until after the com Second, the Authority's current analyses, together wtth the unique circums features of the FitzPatnck plant, do not justify installation of a hardened wetwi TW secuence. The Genene latter inappropriately prescribes a generic modficatj ceciotoly plant speerfic severe accident issue.

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Third, SECY49 017 (upon which Generic Latter 8916 is based) contradicts both itl and other NRC sponsored stuc es on several technical points.

T paoer covelcos these and other issues related to severe accidents and containment l

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Page 3

li. BASE'S W4Ws potentd Mark I enhancements are dit0L200 Clin SECY4417. As Qenene L.stier 0614 (meterence 1), all SECYe.017 Mert I enhancement

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' hardened vent, are to be evaluated as part of the IPE/PRA program. These o L

enhancements are to be subsumed into the IPE/PRA prosess because a plant specif i-analysis must be performed to account for posseble design Offlerances among vanous nuclear power plants that may hear on the necesstry and nature of specific safety imoroveman However, as stated in NRC NUREGs, the une of a hardened vent also requires a Sp '-'*sfic analysis. Applying the same logic leads to the conclusion tha haroenec - p puld also be examined after the IPEs are completed. There distinctive ateut the hardened vent, compared to other suggested enhancements, th Justifies a different course of action.

This recommendation is consistent with earlier ACR$ recommendations, conclu drawn by the CRGR (Reference 5), and Addendum i of NUREG/CR 5226 'An Overv SWR Mark l Containnent Venting Risk implications,' (Reference 6), This NRC re concluded in rts Executive Summary:

j 4

' Applying the Peach Bottom resutts to other plants requires careful consideration. The sequence frecuencies, local peculation censity, evacuation plans, and unique plant features could have a t;;nMeant impact on the resutts?

i Therefcre, the ha'cened vent, like other Mark i enhancements, requires plant and site I specific considerations and should be treated in a consistent manner; i.e., folded into the IPE/PRA procats Generic t.etter 8916 did not establish the need to move with urgency to install a harcenec vent. In fact, SECY 86 206 ( Aeference 7), reaffirmed that the risks from BWR Mar I's are now. More recently, the June,1989 issue of NUREG 1150,' Severe Accioent Risks:

An Assessment for Ave U. S. Nuclear Power Plants' (Reference 8), displayed very low for rts reference Mark i plant, Peach Bottom. These low risks were achieved in spite of Peach Bottom's relatively high 50 mile population and without dependence on a haroened vent.

Origmany, the Commission set out to establish a more balanced approach between acetcem prevention and mitigation. This philosopny was the regulatory underpinning for the Contamment Pe'formance Imptovement Program. Sance 8WRs already have low core mett frecuencies, emphasis has been placed upon reducing the likelihood of drywell shell failure 1 assuming core mett conditions had occurred, i.e. mitigation. Drywell shell failure is ioentiec in NUREG 1150 as the onncipio contenment falute mode in Mark I plants. Yet, g:ven a core mett situation, venting is ineffectsve in reduceng the probability of shell failure.

Consecuently, the harooned vent strategy put forth in Generic L,etter 8916 is inconsistent with the origmal purpose of the Containment Performance improvement Program.

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t The actual risk reduchon potential of a hardened vent pnmer#y Wequency of the TW sequence (lees oflong term decay hast

@ range of TW esquence Wequences was esplayed. The Pit Pat hequency wiil not be known until res IPE/PRA is comp 6 sed. Theref However, there are numerous inchcatens that it vnli no Many of the benents claimed for the hardened vent depend upon the o an afternative source of water, venting in some ai l

injection and acce!erate core damage Successful vent operation, therefo i

L consiceration of other plant specific features beyond the hardened went itsoff lower accioent frequenc:ss or severrty. To desitoduction valu knowing the basefine TW sequence frequency, gn a hardened vent early wrthout e is urjustifled. A hardened vent designed 1

without knowledge of other plant systems is poor practice. Knowledge of t i

reduction of other enhancements should be a prerequisite.

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..2.

4 C. REVIEW OF SECYe@17 l

The recommendation 1e told the herdened vert decimen into the M/

also based, in part, on our remw of SECYect7, the Mark l Cortenment imp Program. Our review concluded that the technical and econome just$cate for haraoned vent at the James A. Fit:Petrek plant was not provided in that documem.

Our review of SECY SM17 raises questions in two broad categones:

Cost /beneftt analyses anc artomatives to harconed voms. These categories are examined below, i

ill.A Cost / Benefit Analyses lilA1 Basic Equellon The basic ecuation used to calculate the cost /benent ratio in SECYe 017 is in e

)

SECY 4M17 proviets the following formula:

i Cost Benept =

A '"#

(Insahnon Cost. Asened Orune Cosa) l Shov1C ins *fla: On Cost scual the Averrec Ons#e Costs, the Cost Senefn, by woul0 become infinite. Even when these two factors are not ourte equal, the Cest sene's ratio is criven by how C10se they are Ed b9Comes insensttive to the Averted EQosure or i

the absolute mag %tuces of the lastalleton Cost and the Averted Onsno Costs. Ther this equation is imprcperly formulated.

This ecuati0n has been recently reviewed by the Electric Power Research Institute. EPR:,

in its era *t report, NSAC 143, *Questionsole Technioves Used in Cost Benefit Anal Nuclear Safeg Enhancements,' Septemoor 1989, (Reference 9) and found to be '

t ina:oropr. ate, lilA2 Present Worth Analyses information from the same EpRI era *t report indicates that the beneet portion of SECY.

SM1Ts haa"? eMeets cost / benefit ma!ys:s may have been overestimated by about a fato of 2.5. S:nce nea'tn eMeets cominate SECY 4H1Ts benefit Agures,it is crucial to calculate th:s f,gure correctiy.

The EPRI cocument advances the argument that any future expense, including future hea'in eMeet costs, must account for the time value of money. Note that the EPRI analysis.;

coes not discount heatth effect costs themselves. Rather, such costs are treated as a constant potential expense that may occur at any time wrthin the assumed remaining 25 yars of plant life Stancard accounting practice for a potential Emed expense yields a time we ;Mec p'esent worth factor of 2.5 for a 25 year period and a 10% discount rate.

Page 6 l

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Even though heWth sNeots egets are harneelves nel de0Durted with e

tooey to hanee a potwt0 Mure expense W any lens must utilize preser v

A appears that the SECY to#nent ed not asequatWy assourt ter th s, _,

and, therefore, overestameles' health eNect benents by about a lector W2.8.

lilA3 TW Sequence Person 4em Figures of about 3 to 4. Using the person rom Agures on page 1330 210 = 1120 person roma/RY assoc 6sted with TW teQuences at Peach B such sequences have a frequency of 10'*/RY, DMeing this person tem number treguency, one obtains a value of 1.12x10 person rems / release. However 7

terms developed in BMI.2104 (Reference 10) for TW seguences and matching terms to 3.82x10 ptson rom data for Peach Bottom from NUREQ,11505, yields 2.87x10' to person rems. Thus, earlier analyses of the Peach Bottom arte wfth source terms associatec wrth TW sequences result in fewer person rems by about a factor of 3 to 4 Similarly, the Peach Bottom person roms for ATWS sequences appear to be abou R

factor of 4 too high in this SECY.

lilA4 PopulatJon Ettects SECY BM17 does not properly account for $tte to stte population d.tferences whien directly sNect the number of person. roms per release. For exampe, the-50 mile radius populaton for Peach Bottom is about 4.1x10 peope. At the Frt2 Patrick plant, the 50 mile 8

radius poputabon is only about 0 64x10' people or about one fttth of the 50 mile populatio of the Peach Bottom site. Basec on information in NUREG/CR.2230,' Technical Guica for Siting Criteria Development' (Reference 11), this smalier population at Fit 2 Patric reduce the Frt2 Patrick person rems /TW sequence release to about 40% of the Peach Bottom figure. Page 21 of Enclosure 4 (Regulatory Analysis) of SECY 49017 discuuss effect of population ciftetences on risk, it is acknowledged that leta populated arte resurt in risk rMuctions of 4Dogi a factor of five. The SECY then accreases the poss that the lower station blackout (SBO) frequencies at Peach Bottom my offset that plants' higner population. The SECY also c4 cusses potentaa! circumsta% Wert both highe population anc hign 580 frequencies may coexist. However, the MCY is silent on combinations where both SBO froovencies and population eensiten are low and therefore -

)

the cost oenefit rat o would be below the SECY generated numoors. Yet, subsect to the Authorrys iPE/ PRA resutts, this is likely to be the Frt2 Patrick situation.

l 1/ See Peach Bottom source terms bins, Volume 2. Appendix E, NUREG 1150, eratt for t

L comment, February, 1987. 2.87x10s

• niie 3 62x10 8 person. rems corresponds to a CRAC 2 calculation L

person. rems is a MACCS calcutstion. The BMI source term is composed of 0 046!. 0 045Cs, anc 0.15 TE.

Page 7 l

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NIA3 Net Risk Analysee To comprehensively ecjuste the coet/benent ratio of a herdened went, one must consider both the potenbal beneftts and the estrimeme of this plant moesncaten. Esamples of this logic can be seen in Tables 2,3 of NUMEQ/CR435, snached. Note that venting can both increase and secrease noks. In actual piam apoorne analyses, numenaal values woute replace the greater than and less than symbols, esquence by. sequence These numencal values wovid be weighted by trer sequence frequencess and then summed to determine the not nsk. These numencal values would also be sPfected by the particular vent design s

~

ano operaton. Por enample, eaHy ventmg may result in releasing a tarper amount of noble gases than celayed venting. SECY464t7 coes not partcularly escrees not risks, nor could rt. Such an evaluation requires a speerfic plant analysis.

Pan of a not risk ana'ysis includes an accountin0 of the asposure of plant personnel while instaning plant modtfications, it does not appear that exposure of plant personnel was accounted for in the SECY's cost / benefit analys:s. Much of present plant personnel exposure occurs dunng outages and maintenance Using national SWR exposure values,if piant personnel were exposed to a one time ten percent increase over normal exposure levels because of installing a haroemec vent, this woulc be $t person. rems (See Mgure 1, ' Collective Radiation Exposure

. BWR Ave' age? Using rtzpatrick specfic sata, a comparable figure would be r

l apprcrimate!y to person rems ) If the frecuency of the sequence that this flx was to mod;fy was inttially 10 s/ Y and twenty five years of plant Irfe remained, this onsite exposure is R

ocurva'ea.t to abo / 290.000 person. rems of offsite exposure, at a plant capocrty factor of 0.70. Theretore. SECY 8S417 is incomplete without an analysis of the onstte exposure due 8

to insta$ng these proposed modifications. If 1.4x10 person roms approximates a TW r

see,vence releast at the Ft:2 Patrick site, onstte exposure would be equal to about 20% of this figure, based on the above assumed exposure rates and sequence frequency, Part of the justification ot'erec for the hardened vent is the SECY49017 argument that venting lowe's the containment backptsasure on the ADS vatves. With lower containment backpressurt. the ADS can tnen open, which would lower the reactor pressure. This,in turn. covic alio* a oiesel oriven fire pump to deliver acequate amounts of water to cool the core. This neovence of events may be appropriate for certain types of ADS designs, but it may not be universally true. Ma k i ADS valves fan into a number of categories, e.g. two I

stage vanes, three stage va!ves and eitCtromatic valves. The relationsnip between conta;nment pressure and ADS valve operability cepends on the type of valve. Por three stage vafves, as was appa'ently used in SECY4E 017, once the valve closes the increasing reactor sy stem pressure will oppose reopening amj only a large enough escrease in contanment pre &sure will cause the valve to reopen. Wrth such a configuration, venting wouic De use+ulin opening the va!ve. With a two stage valve, once the ADS valve closes it can reopen rf either the reactor pressure increases or rf the containment pressure cecreases, Consequently, containment venting is not required to open two stage valves.

8

- 2/ The figure of 1.4x t0 person roms is denved from using the Peach town TW 8

secuence MA005 vasue of 3.62x10 person roms and correcting for the sma,'er FitzPatrick popviaton 1

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i Therefore, the hardened vent 'oreeft' given in SSCYe017 ter assiebng the A i

would only be applicable to that subset of Mark I pierts that have three stage van i

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mitPetrick uses two stage AD6 vanes. Furthermore, N there are eseel e water delNory conooms, these can be addraesed in a venery of ways auch as booster pump, changing over to two stage ADS veNes, etc. This potenhal pum conoom does not maneste the use of herosned vems for as resolution.

Pan of the crodrt SaICY 4D 017 assi croort may be unjustrfice. Por many A e hereened verns la for ATWS esquences. Such auwences, te vom flow area required to prevent containment over pressunzation would M much larger than one used to cope wfth just occay heat as in a TW sequence it appears that the Boston Reson design referen by Genenc L.atter 8616 would be limrted to TW type sequences. A vent capa ATWS secuences would be considerably larger and more esponeNe that the estimat teferred to in SECY 49017 and Generic L,stter Sig. Therefore, there seems to be en inconsistency in the ATWS nsk reduction credft gNon to herdened vents and the as designs and costs. Futhermore, application of the ATWS rule has micumized the risk reouction potential for vents for such sequences.

There are other complications between ATWS sequences and venting that needs to be resolved. If a TW based vent design does not provide sufficient containment pressure relief dunng cenavn ATWS events anc 11 venting has Imle impact on preventing drywell theit attack upon reactor vessel failure, the r:sk reduction potentialin ATWS sequences is curte limited.

On trw other hanc. concems have been raised that rf these plants were vented cunn secuences, core damage could occur because of induced NPSH problems. Previous ana.'yses of cenain ATWS secuences show that venting might induos a core mort artuation.

inoeoencent ana'yses of the Brown's Ferry and MtzPatrick plants show that certain ATWS seovences go through a penod of power oscillation and then settle down to thermal equihteium, wnhout oore camage. Both of these plants use two stage ADS valves, Other plants wrth three stage ADS va'ves in simdar ATWS events are predicted to lead to core camage,indecencent of vent operation. SECY 86017 does not discriminate between piants wrth eitterent ADS valve configurations and, thereby, can increase nsks due to cena;n ATWS evems for two stage ADS valve plants. This example underscores the importance of inoivicua' plant examinations anc the jeopardy of imposing genene flaes.

l The inrtistion of voming is symptom based in an ATWS sequence, high containment I

pressures, enougn to inrtiate venting, could occur much more quickfy than in a TW secuence because of the much greater energy generated in ATWS, Present EPGs (Emergency Proceeure Guidehnes) for venting Mark I's minimize the likekhood of NPSH l-proo ems, but s's largely based on TW sequences. The adequacy of these EPGs to prevent 7

l NDSW prc:* ems in ATWS seovences may have to be reviewed. This infers that application of ventng to ATWS sequences may require considerable study prior to starting any otsign or cor'struction ettort.

SECY 89017 identMas a number of benefits to ADS operation, such as lowering the hkekhood of prompt containment failure upon reactor vessel failure. One maior ADS benefit was overlooked. Operation of the ADS pnor to reactor vessel failure has been investigated for Peach Bottom station blackout secuences. ADS operation significantly coals the reactor i

intoma: suria:es and this.in tum, resurts in the volatile radioru,lices released from the fuel being trapoec on these cooler surfaces. One sucn Authorrty analysis of the Peach Bottom piant showoc aoout a factor of 10 reduction in the source term when the ADS was used. To Page 9 i

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N entent that ADS opecion toduces the source term in venous esquence teduchon potertial of other plant meencebens, such as vems, is redw not annoum ter this adetional ADS benem and, therefore, overreed the i

vents.

benents of harsened vems is a recuoec likelihood at reactor vessel fanure, proviend that voming is 6rvtese earty in the esadem seq the contenment pressure spike caused by a reactor ve ADS could accomphsh the uma thing. The comen reactor vessel failure is reduced because of the earlier energy transfer from th; the suppression pool via the ADS. ADS operation is like 'intemel voming,' l.e lower peak contenment pressures but wfthout the earty release of noble gas Th. fore. ff a vimey e.cida to mwmae tne um of = ADe in ns wvere.co,dem management program, the benehts of harcened venting are decreased. However, thei process outhnec in Generi: Latter 8916, which separates the venting raave from other potent al Ma'k i enhancements, complicates matters. This separation process leads overestimates of the worth of venting.

l Perhaps the mest er,reme example of overlapping between the hardened v potentra! plant modications is the impe:t of contenment sprays. As SECbekO17 out, sprays woule recuee the trborno concentration of fission products in the cry woute retard or prevent core concrete interactions and would reduce the likelihood of Crywell shell failure. Sprays may also reduce comainment pressure prior to reactor failure,if initiatec saHy enough. This woute,like venting and ADS operation, ree probabiirty of a prompt conta nment overpressure. As discutted later, sprays could used in TW sequences to ace wstor from other tources to the suppression pool via 1

Cowncomers, thereby artending the time for recovery because of the additional heat capacrty. Therefore, sprays accomphth many of the same risk reduction beneftts of v anc some (e g., reoucee probabilrty of cywell theit failure) that venting does not. SECY.

89 01rs person. rem analyses cid not account for the spray system's abiltty to prev ane/or mitigate many severe acoceats, espally rf backed up Dy e ciesel enven pump. 1 By ignoring tne potentia) benefits of the comainment sprays, the nok reduction po vents is overratec.

lil.A,6 Summary The process of accelerating hardened venting in eevance of other plant enhanceme lea:s to cuerrating the nsk reevetion potential of venting. The process of designing v pnor to IPE/PRA completion can lead to inferior resutts Finally, Genene Latter abi6 cannot be properly impiomented since acceent frequencies would be unknown. Therefo the NRC cannot previce a plant speerfic cost /beneht analysis of hardened vents without completion of a plam speerfic IPEiPRA.

l.

+

L t

Page 10 L

L

- - - --- ^.

_.--.--.---------^-~~~^'1 N*S 1

M D NM MD 0.5.1 OveMow 1

IM 1 hit tedtion themetkes to hardened verts we q Agures proWIed in SECYect ?3 TW esquences rep $ecuated. Saeed on the pers resort about M to 48% of the nok reductson potemist for the vanous acedent escuences that were enemmed. S there are a number of ways to reduce the nek pote of the hardened vent, if attomatives to a herdened vent we elected to reduce TW sequences' risk potential, the residual risks of the ATWS and station blooko and the station blackout rule probably retuft in ev 3

than were assumed in SECYe417, 1

The TW toquences, therefore, are central to justNying a hardened vent, if t tequences have a low base line frequency or il these sequences are minimized l

J attemative means, then hardened vents are not justf6ed, l

loss of containment integrity and core camage can b haroened vent.

l The Authorit/s poston is influenced by the fact that there are many ways to plant. This fact has &!rea:ty manifested itsett by the low calculated core melt fre increases the number of oper4 Die systems relative seovences. TW neovences evolve vo'y slowly, giving the oporstor long periods of tim!

take corrective action. In fact,if TW tecuence piartt recovery cannot be accomplishe i

BWR wrth all of rts heat removal pathways, wfth electric power available, slowly evolving aCCicerft, then serious Questions ante about the overall Value of op currog severe accioent concitions and the usefulness of the whole severe accident management progeam.

in the very unlikely event that a TW 6ecuence does lead to a core mett, the relaatei radioactive material to the environment could be minimizac by corttainmortt sprays.

Because both a low probability of havmg a TW necuence core mett is expected, especiary when anernatives to a hareenec vent are used, and very small source terms o

should resurt rf containment sprays are operatec, Fit 2Patnck TW tequences are not l

.xpect.e te be nsk signeicam..

I l

3/ Page 10,5ECYMC17 l

l 1

)

Page 11

.i s

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ill.3J First Steps Well be%re herdened vents would be eenedered et MitPetnok, severW taken. The first step is to determne what the TW escuence Wequensee ere. e r.

now in mobon es part of the PltrPetnck IPE/PRA. Assumng that the TW esque freousne.as are too high, a number of addebonal eteps may be taken to mnmis is onecnDoc in the oreft EPRI report,' Severs Accident Concenser Isolations,' EPRI RP 242>22, May 19N (Reference it). This report i review at the Authortry to assess its usefulness to the Mt1Patnck plant.

i 1

til.B.3 APtemstlve Means 1

if, after applying the suggestions in the above EPRI report, M was determin i

secuence frequencies wera still too high, the use of afternsthre means to establish a 1

pathway to an ultimate heat sink would be investigated. Several utilities are i examining such afternatives and a number of member utiitties of the SWROG have sponsored studies by the Genera! Electnc Company on such anomatives.

Uncer investigation at this time a.'s:

i (1) Opening of the main stoa.m itolation vatves, I

(2) Removing heat through the main steam isolation valve drain lines, (3) Removing heat through the reactor water cleanup system, (4) Rooding the wetwell torus up to the drywell/wetwell vacuum breakers, and (5) Combinations of the above.

Item (4) rioes not establish a pathway to an uttimais heat sink, but the additional heat capact er.encs the time to recover from this accioent prior to core damage. One way o increasing the suppression pool's heat capacrty is to operate the drywell sprays. This would also recuce the source term, should there be a core mort, minimize core / concrete nteractions, and recues the likehhood of drywo: shell fa# lure.

.j i

j Other a'tematives to a hardened vent not yet under active review are:

(6) Heat removal through the dr)well coolers (7) Mass and heat removal through the torus drain line (8) Mass and heat removal through drains from any system that takes suction from the suppression pool (e.g., the RHR system)

(9) Heat losses through the torus surface itsett, postibly enhanced with fire hose sprays, and, Page 12 b

m A

,_m e.-

. _ _ _ _ ~ _ _ _ _ _

(10) LOW pressure veting (ventmg without emeestbng preneure bou As more utilities tum theit attention to WtemeNes to hertioned wants, cIhe i

may be identmed.

Note that ft is not necessary for a single attemattve to b secuences. Since the comainment pres.sure noe rate is very precual in TW esquences, particularly if the pool's host capactty has been increased, containment pressure eq may be achieved by the sum of two or more eftemattves, o

4 til.B 4 FNrPetrick Specmc Studies The Authority has actively participated in the BWROG/GE effort on these aptomative pathways. Furthermore, the Authority is examining the application of the GE report specificajty to the Fft2 Patrick plant. In)tial reviews of the Rt2 Patrick reactor water cle system and use of the FrtzPatrick MSIV drain lines indicate that heat removal with a combination of these two systems is close to matching a decay heat rate of 1% of full power. The Authority is also in the early stages of examining the mass and heat removat capacilities of the torus cra;n lines, RHR dren lines and other systems that take suction from the suppressich pool.

111.8.5 Summary The Frt2 Patrick IPE/PRA may show that this plant already has a low overall TW sequ treQuency. If not, there may be opponunrties to le*er the frequency of TW sequence initiation and further opportunities to recover trom such sequences, once initisted. Should all this still prove to be inadecuate, a vanety of artomative decay heat removal means.

appear to be feasible for Frt2 Patrick, based on plant speerfic studies now underway, in the vey unlikely event that a TW sequence induced core melt does occur, the Consequences of such an accicent Could be minimi200 by using the contenment spray system. In the event that normal crywell sprays are inoperative, backup diesel driven sprays mignt be used.

in view of all this,it appears that the TW s4Quences for Mt2PatnCk will prove to be an.

unimportant risk contributor and the hardened vent unjustified. In any case, the decision sncJe De mace upon comp letion of the iPE/PRA process.

i 9

k Page 13

-4 a.

-,., +, - -. - -.,

IV. REFERENCES 4

' NRC Generic Latter 514 dated September 1,1W, James G. Po 1.

regarding the same 4Wect.

i 2.

NYPA letter, J. C. Brons to NRC, dated January it, iW (JPN46@1/IPN4M01) repareng reveons to comments en NUREG 1150, i

3.

NYPA letter, J. C. Brons to NRC dated September 3,1987 QPN4761/IPN4 045) regarding comments on eratt NUMEG I tto.

SECY4H17, Mark i Cenment improvement Program, ested January 23 4

5.

NRC January 5,1989 memorandum, Edward L Jorden to V Stello, reg minutes of CRGR Meeting No.152.

6.

Addendum I to NUREG/CR 1225. EGG abas,

• An Overview of BWR Mark 1 Containment Venting Risk implications; An Evaluation of Pctertial Mark 1 Containment improvements.'

7.

SECY 687J6. catec July 15,1988 NUREG 1150. June 1969 itsve, ' Severe Accident Risks: An Assessment for Five 5

8.

S. Nuclear Power P ants.'

i 9.

NSAC 143, 'Ovestionable Technieves Used in Cost Benefit Analyses of Nuclea Safety Enhancements,' Septemoer 1989.

-l 10.

BMI 2104 11.

NUREG/CR 2239, ' Tech',4 cal Guidance for Siting Crrieria Development' 12.

EPRI Re: ort EPRI RF 2420 22 ' Severe Accident Prevention: Reducing BWR

- t Concenter ISolstions,' May 1989.

h

}'

)

,9i t

k l

i L

Page 14 f

t -

V. TASt.gS AND FICyggg i

Tat:11.

Risk CompeMeens et tason Bleekevt Scenados Mtn YsMove Venung Systems from NURE0/CR 8335 Quatttstive Change in Risk Reistve to a Nonvenung SoonsMo NARo PzPE VEN1 $YsTEM DUCT WITH A WITHOUT A WITH VENT $YSTEM

[gETURE DISK RUPTURE DISK FILTra $Ystt*

AccrotNY ENo STATt EARLY L&B

[AMI L&I$

[Any (g3 (ARLY M

VESSEL FAILURE WITH NC NA NC NA NC NA NC NA EARLY LINER MELT-THROUGH

)

VE$SEL FAILURE WITH NA NC NA NA NA EARLY CO' TA!hMENT DVERPRES$URE FAILURE RECOVERY WITHOUT NA NC NA NA NA VESSEL OR CONTAINMENT FAILURE RECOVERY WITH VE5SEL FAILURE AND WITHOUT CONTAINMENT FAILURE VESSEL FAILURE Awe LATE bRYWELL CONTAINMENT FAILURE LEGEso:

EARLY INITIATE VENTING sEFORE YEsstL FAILURE AND LEAVE OPEN. (THE RUPTURE DISK WILL NOT OPEN UNTIL VESSEL FAILURE).

LATE INITIATE YENTING AFTER VE5SEL FAILURE.

VENTING CCNSEQUENCES ARE EXPECTED TV st LESSER THAN(<)

THE NONVENTING CONSEQUENCES.

VENTING CONSEQUENCES ARE EXPECTED T0 SE GREATER THAN (>) THE NO CONSEQUEHCES.

NC NO CHANGE IN RISK.

NA NOT AVAILAsLE, VENTING WOULD NOT.sE INITIATED AT THIS TIME.

Page 15 b

m'

3 Table 2.

Riek Comperleans of ATWBandTWh 14 +

S...

With Various Venting Sm from NURES/8228 QualNedve Change in Risk Reiseve to a Nonven9ng Seeneno 1

HARD P PE' VENT SYSTEM i

DUCT WITH A WITNOUT A WITN F2LTER' PARAMETER VENT $YsTEM.

RuptuRr DrsK-RuptuRt DrsK System 1:

CONTAINMENT-FAILURE VENTING-SCENARIO $

WITN. ALTERNATE 4

4 4

INJECTION

]

. WITHOUT ALTERNATE NC-NC' NC NC tINJECTION AND WITH EARLY LINER-MELT =THROUGH-WITHOUT ALTERNATE

!NJECTION AND WITHOUT

'EARLY LINER MELT-THROUGH NO CONTAINMENT FAILURE

'. VENTING SCENARIOS.

WITH ALTERNATE NC NC NC.

NC INJECTION-WITHOUT ALTERNATE 1

INJECTION ~

LEGEND:

c'

- VENTING CONSEQUENCES ARE EXPECTED TO BE LES$ER THAN (4) THE NONVENTING 3

-CONSEQUENCES.

>l

' ;VEhTING CONSEQUENCE $'ARE EXPECTED TO BE GREATER THA4 (>)-THE MONVENTING i

CONSEQUENCES.

NC: - N3 CHANGE:IN RISK.

Page 16 i

71

..j i

Collective Radiation Exposure l

BWR Average Man-rem per Unit-year 1,400 1.230 1,200 1.137 I00l y

1.017 1.003 j

1,000-

{

800 i

600 622 521 511 i

400 0

1980 1981 1982 1983-1984 1985 1986 1987 1988 1990 Goal J

-.g a

s w.m--

a

.sm.1...a.

a---.a

• 1 to JPNG.070 j

i

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1 s n,s i

New York Power Authortty Cost Estimate for the installation of a Hardened Vent et the James A. Pit:Petrick Nuclear Power Plant introduction 4

A prelir3inary estimate of the installation costs for a hardened vent described in G 8916 and GECYW 017 bas been preparedc Secause of the short amount of time allotte prepare the estimate. h is not comprehensive and therefore should not be considered final.

{

a Accordingly, the Authorrty reserves the right to revise this estimate, Conceptual Design This estimate vcas casec or, a conceptual design similar to the design outlined in the Letter anc implementec by the Boston Ectson Company at Pilgrim. It assumes that structures, systems and components will be used to the maximum extent possible.

i The existing containment Vent and Purge penetrations were assumed to be the connec the conta.nment atmosphere. Piping from the torus penetration to the SBGT (Stand B Treatment) ritter train and from the fater train to the plant stack is ameady *hard' pip cesign pressure of 150 psig. The SBGT fitter trains and transition p'eces (located outside of.

l seconday containment in the SBGT room) are 'sott' ductwork with a design. pressure of approximate!y 1 psig.

l Essent.a? feat.;res of the conceptual design include:

New 12 inch, seismic class I, 350 psig, piping and suppods a

necessa y to bypass SB3T f:lter trains. (A.12 inch nominal pipe size was selectec to match the larger of the two limiting pipe sections upstream of the SBGT fitters.

A new isolation va've and rupture disk to isolate the $8CT bypass trom the SBGT.

Modifications to existing valves (including containment isolation valves, and SBGT valves) to assure operaDility at severe accident pressures and temperatures.

t.

SBGT piping support modifications to assure openipikty at severe accicent prassures and temperature.-

s Additional controls and cor'trollogic modifications to operato how sospment - Controis were assumed to be located in the main control room.

j 1

g i

i..

~

n-

.-,....- - - ~

- Attachmem Il te JPNha70.

Addition of a now, hgher range plant stack resation monitor to detect the release.

f j

-. AC power to new electrical soupmert.

'i DC power to new and existing equipment for the AC independent

~

option.

o..

- Costs included in this estimate' include: engineering, design, material, and ins base estimate (non AC incepenoont option) assurred the avaHability of AC po include a 25% contingency.

j This estimate does not consider costs associated wtth redistion exposure, hea support, anticontamination clothing, licensing costs for techrkcal specrfication chan

, rewrrting operating procedures or staff training.- Additional costs and unforeseen problems ca q

- only increase the total cost, The estimate also does not include replacement power cos u

mod 44 tion is tne on the critical path for an outage, an increase of $500,000.00 per additio outage day due to lost revenve/ replacement power costs would be added,

)

Results..

\\

Base Estimate = $680 000.00 1

incremental cost for.

AC incepenoent cesign = 8 70,000.00 6

(

2-e si

. q

@ v.a ;

i

'l N UTIUTIES-o.a.. em

.s.

asi i n.~a ceaa.eice l

' T ; m:'.r.rt,t=

,,,anfrom c'owwectcu' os tabca?O

' ) ) ***'

l October 30, 1989 i

E 4

Docket No. 50 245 AD.ini Re: Generic Letter 8915 o

ISAP Topic 1.113 m

U.S. Nuclear Regulatory Commission

'y' Attention: Document Control Desk l

Washington, DC 20555 Gentlemen:

o L

Millstone Nuclear Power Station, Unit No. 1 Response to Generic Letter 89 16 L

Installation of a Hardened Wetwell Vent

/

Durrese and-Introduction The pur;ose of this letter is to outline the approach'that -Northeast Nuclear Energy Company (NNECO) intenes to utilize in resolving the Mark 1.contain e.t narcened vent patn issue at Millstone Unit No. 1.

This letter is in response:

10 Gereric Letter 8916, " Installation of a Hardened Wetwell Vent," receive:

on Se:te :er 14. 1989, 3

• • e Staff icentified a number of plant _ modifications at' the conclusion of-the Marx -1 Containment improvement Program that could potentially enhance existing

a:a:ti);y to: prevent and mitig' te the consequences of severe accidents, e

As-part cf the comprenensive plan for resolving severe accident issues, the-CO- "s51on' concluded that the recommended safety improvements with:one exce:-

-ticr. har ene:

wetwell vent capability, should be evaluated as part of the In: 1vi.ai Plant Examination (IPE) program.

To address the hardened wetwell-

^

vert. Generic Letter 89 16 was issued.

It requested.that licensees of plants niin Mark I containments provide the NRC Staff with plans: for addressing the 1ssue of hardened wetwell venting capability.

Encouragement was also provicec to voluntarily undertake plant modifications under the provisions of 10CFR50.59. Absent voluntarily incorporating design changes, it was requestec that cost estimates for implementation of'a hardened wetwell vent, described in SECY 89 017 be' provided, including.an incremental' cost estimate.for installation of an AC independent design'.

These cost estimates. - along with the-Staff's plant specific backfit analysis, will be used to evaluate the efficacy of recuiring the installation of the hardened vents.

intecrated Safety assessment procram NNECO acknowle:ges that there may be a potential benefit' from installation of a narcen6: vent path, but we co not see this as an issue that is best treate:

cuar: at...as t

1

(

t

J

.U.5, Nuclear Regulatory Coenission

.A08591/Page-2 October 30, 1909-independent of our IPE program and addressed in isolation. This is especially true since our front end evaluation is essentially completed and the back end L

is tscheculed for completion approximately one year from now.

Given the_ need to' prioritize plant modifications, the Integrated Safety 1

Assessment Program -(ISAP) has become an integral part of project evaluation for Millstone-Unit No. 1.

Our experiences to-date have demonstrated the a

program to be a resource efficient and cost effective process for enhancing

- the safety of Millstone Unit No.1, while also being esponsive to long-standing NRC support of systematic safety reviews of operating nuclear power 1

plants.

Moreover, NNECO's comitment to the. "Living Probabilistic R1.sk Assessment (PRA)" concept contributes significantly to the overall safety improvement process.

l "Livino Drebabilistit Risk Assessment"

~

(

J Northeast Utilities maintains'a corporate-policy on nuclear safety goals.

It is cur-intention to implement this policy through the "Living PRA" concept.

' As needed, the Millstone Unit. No.1 Probabilistic Safetyi Study (PSS),-a.

~

1 L

Level 1 PRA, is updated to reflect design and-oper i.nitial submittal' to the -Staff on July 10,1985,gt)ional changesa Singg two major updates have been submitted.

NNECO recognizes the importance of the Millstone Unit No. 1 PSS to understand-u l-ing the characteristics of the plant and to the reduction of risk.- In con-3 junction witn ISAP, the PSS is an important tool for prioritizing the expendi-ture of resources in a way that will'be most effective in reducing the overall

= risk 'to. the public.

As a direct result of the PSS analysis' the Millstone Unit' No, 1 Core Melt ~ Frecuency (CMF) ~has been reduced significantly, 'to 8,92 x 10 5/ year, through improvements, corrective actions, reanalyses, and Lprocecural.mocifications.

l F:r.Millstere Unit' No.1, a Level:1 PRA' for internally initiated events, as well as _ fire anc internal flooding, is complete and actively exercises.

As i

4 I

(1) 'J.

F. Opeka' letter to J. A. Zwolinksi,

" Millstone Unit No.1

. Probabilistic Safety Study Results and Sumary Report," dated July-10, L

1955.

p (2)

E. J. Mroczka letter to U.S.

Nuclear Regulatory Comission,

• Probabilistic Safety Study Update," dated February 11, 1987.

"(3)' E. J. Mroczka letter to U.S.

Nuclear Regulatory Comission, " Millstone Nuclear Power Station, Unit No. 1 Probabilistic Safety Study Update (Revision 2)," cated February 10, 1989.

y

[s) 1

~

' U.S. Nuclear-Regulatory Comission

,r A08591/Page 3-'

October 30, 1989 I

discuss'ed in our July. 31, 1986 'I'and' August 4, 1987(5) submittels, NNECO has I

plans in place to expand the-Level 1 PRA model to. evaluate containment

. response.

This - ongoing be nd analysis effort was further discussed in a-

letter datec July 27, 1989 the IPE program guidance-of Generic Letter 88 20.in which we described our ai Upon completion of-the s

back end analysis,- currently scheduled for late 1990, NNECO believes - the, safety benefits of implementing the Staff's recommendations on Mark I contain-eents, including hardened venting capability, will best be ascertained.

Mack 1 Containment Princieal-Accident Secuences Generic PRA studies for boiling water reactors (8WRs) indicate that 'accide'nts initiated by transients rather than loss of coolant accidents (L 1

dominate the total CMF' estimates.

The principal accident sequences consist o station 1

blackout (550), anticipated transient without scram (ATWS), 'and -loss of

[l loa.g terS decay heat removal- (TW).

NNECO's progress towards resolving these-L 1sbes is discussed'below.

o Station Blackout ISAP Topic 1.106 was assigned to track-industry initiatives relatingi to'-

SBO, as-well as the in house effort to assure that Millstone Unit No.1.

complies with the' SB0 rule.

This topic will also -incorporate spec'fic open items thattresult from the NRC Safety Evaluation for the Millstere.

I sitt visit of July 18 21, 1989.

Several emergency gas turbine generator (GTG) reliability improvement mocifications were implemented..during the 1989 refueling outage under other ISAP topic numbers.

They are related to the-topic of-SBO, and were discussec fully in Atte.nment I response letter Edatec April.17, to thj7Irecently transmitted SB0 rule 1989.

Should - the GTG and diesel generator be unavailable, the 23 kV Flanders line can supply power to tne

. c.

(4)

J. F. Opeka letter to C. I. Grimes,

' Integrated Safety Assessment

(

. Program Final Report for Millstone Unit No. 1," dated July 31, 1966.

(5)

E. J. Mroczka letter to U.S. Nuclear Regulatory Commission," Integrated

' I Safety Assessment Progrssi (ISAP)," dated August 4,1987.

(6)

E. J. Mroczka letter to U.S. Nuclear Regulatory Comission, " Response to:

1 Generic Letter 88 20 -Individual Plant Examinations for Severe Accident

/

Vulnerabilities," dated July 27, 1989.

l (7)

E. J. Mroczka letter to Dr. T. E. Murley, " Response to Station Blackout LI h

Rule," cated April 17, 1989.

4 L

4 b

4

~

_ - ~ -

4

U.S.LNuclear! Regulatory Consission~

A08591/Page 4'

i October ' 30, 1989-4 3o emergency station service transformer, which feeds emergency loads through Bus 14G.

1 for reducing risk of severe accidents resulting from 580 A 4 kV cross tie exists.between Millstone Unit Nos. I and 2, which. enables Millstone

-Unit No=.,1 to be powered by either of two Millstone Unit No.;2 emergenc diesel generators.

It is available'within I hour of the _ onset of the SB0 event - and has sufficient capacity and capability' to operatez systems-necessary for coping with the event for the required SB0 duration of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, to bring and maintain the plant in safe shutdown, The Millstone Unit No. _2 emergency AC power source can be credited as an AAC : source since it satisfies the Appendix B criteria of NUMARC 67 00 and is avail-able within.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The Staff has already approved a 4ehour duration 1

order to bring the AAC power source on line for Appendix R compliance,g-lf a decision is made to voluntarily install an AC-independent hardened-a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SBO, but rather will design the vent to be con

[

n

{t implementation of the' requirements of the 5B0 rule.

It should be. noted that at this time, no; analysis exists which shows that.an AC independent vent wi.11 be bene.ficirl for Millstone Unit.No. 1.

. The NRC Staff ~ states that one goal of the 560 rule is to maintain the

.. J 1

frequency _ of' core damage from SB0 near cr below 10 5/ year.

In the

- suppl ement a ry information to the final rule, the NRC ~ Staff states 'that the 550 rule must - be met regardless of whether a plant specific PRA currently meets this goal.

The NRC Staff does not, on -the other hanc.

precluce the licensee from identifying plant specific PRA data to support

a. cetermination-that 5B0 would have an acceptably ~ small probability for-causing core damage.

Ace'ordingly, NNECO reiterates..it's ' previous determi-

' nation that the-- CMF of SB0 at Millstone. Unit No.1,.from internally-initiated events,at power,1s approximately 10 5 per reactor year.

t b

c

-f.-tiescated Transient-Without Scram LIs Tne ATWS rule requires that BWRs1 install an-Alternate Rod.' injection (ARI) s y s t em ~,.a Standby Liquid Control System (SLCS) with 4 flow capacity ecuivalent to 86 gallons per minute of 13 weight percent-sodium pentabo-rate solution,, and include features to automatically trip the reactor y

coolant recircuiating pumps under conditions indicative of ATWS.

c (8)

M. L. Boyle lettar to E. J. Mroczka, " Safety Evaluation of the Post Fire

?

b Safe Shutdown-Methodology Yillstone Nuclear Power Station, Unit No.1 (TAC No. 60188)," dated April 14, 1988.

b

+.

s U

U.S.. Nuclear Regulatory Connission A08591/Page:5' y

October!30.-1989 NNECO's. implementation of ATWS is complete, The Technical Specification changes associated wig the SLCS were approved by the Staff in a letter

~

1987.

The Safety Evaluation for the ARI system and dated July.30, automatic 'recir pump trip was provided in a letter - dated

~

October 6,1988ggtingThis completed the Staff's review of Millstone Unit -

No. I's implementation ~of the ATWS rule and concluded that Millstone Unit No. 1 is in compliance with the rule, o -

Lone Term Decay Heat Removal ISAP Topic 2.28 was initiated to-study the long term cooling capability i

at Millstone Unit No. 1.

The study was completed in August 1987 and the results - have been factored into the Millstone Unit No.1 PSS.

The revised PSS indicates that f ailure of long term decay heat removal is no longer the major contributor to total CMF it once was considered to be.

' At Millstone Unit No.1, the TW sequence requires a combination of loss.

of the main condenser, loss of the isolation. condenser (IC), loss of shutdown Cooling, and loss of alternate shutdown cooling incl' ding loss u

of torus cooling.

The~ availability _ of these systems -is affected by' the accident scenario as well as the availability of support systems such as

~

AC power and. emergency service water.

The redundancy' in decay heat re. oval systems make the frequency of 1d sequences at s Millstone Unit No. 1 1ow, approximately 10 5/ year.

Our. review of SECY 89 017 indicates that the-primary benefit of' the hardened ~ vent is the further reduction of risk associated with the T.

.t secuence. NNECO believes this goal mayLbe accomplished by the IC already installed at Millstone Unit No.1..

The IC system removes heat from the 4

core via ' a natural. circulation cooling-process.

As the-steam is con-densed within the !C, heat is. rejected to the shell of the IC.

The condersate flows back to 'the reactor - vessel by gravity,. thereby.conserv-

~

ing tne reactor coolant inventory.

There is sufficient inventory of' wate* on -the shell side of the IC to remove decay. heat for approximately i

E" 45.1rutes.

The IC is station = AC independent.

The only ~ valve which neecs to change state for successful IC-initiation is DC powered.

y 3

Long-term makeup to the IC is -provided by the Fire Protection System-which consists-of three redundant pumps, one-of which is diesel powered r

a ' second which is AC powered' from Millstone Unit-No. 2.

The IC L

o (9)

C. 0. Thomas letter to E. J. Mroczka,-

" Standby Liquid Control System License Amendment No. 5 " dated July 30, 1987.

(10) M. L. Boyle-letter to - E. J. Mroczke,

" Safety Evaluation on ATWS Rule 10CFR50.62'(TAC No. 68494)," cated October 6, 1988.

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j U.S. Nuclear Regulatory Commission

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A08591/Page 6-October 30.-1989 b

makeup system also consists-of a DC powered motor operated valve (MOV),

along with check-valves and locked open manual isolation valves.

An alternate means of IC makeup is provided by the Condensate Transfer System.

In-addition, ISAP Topic 1.02, Tornado Missile Protection, addressed the need to provide 4-missile, protected source of makeup water

-to-the IC.

IC makeup. from the city water supply.NNECO concluded that th Modifications are scheduled - for -

implementation in 1990.

~,

The IC and11ts makeup system significantly reduce the' contribution of TW-type sequences to CMF.

One of the dominant contributors to the_less'of IC during an. SB0 is the loss of the-diesel driven fire pump.

Since the t,ardened vent e

is of little benefit without, the fire pump for makeup' to the reactor vessel, a hardened vent will not'likely have a si effect on CMF reduction for 580 sequences at Millstone Unit No. gnificant.

1.

E e*:ency Oceratino Procedures NNECO imple ented Revision 4 of the BWR Owners Group (BWRO'G) Emergency P dure ' Guidelines (EPGs) through ~a ~ revision to the-Millstone -Unit No.1 Emer-gency Operatsng Procedures (EOPs) on September 1,1989.

The ' generic _ BWROG guidance concerning when' to vent _ containment was followed.

This apptcacr allcos venting before containment pressure reaches the primaryLeontaineer pre s s ure..l_imi t.(PCPL) of-. 71 psig at. torus bottom - or hydrogen and -exyger concentrations exceed 6 and-5 percent, respectively. --The intent is to.take

~

s i

action to reduce the challenge to containment before an uncontrolledicontain-rer.t boundary failure occurs.-

5 The 'EPGs. however, de'not provide retn00 is to vent from the wetwell. guidance on how to vent.

Our < preferred if possible._ If ventin possible, due to failure of the torus, vent-valves g from the wetwell'

'3 is not

for example.- the dry. ell will be vented. ' Based on our evaluation-of several options, we have 4

Oe.eloce:

L a method 'for venting based upon radiation levels in the drywell, if L

cry ei' ra:iation' levels are less than' 40,000 R/hr, primary containment will Tne escacity of the MRB exhaust fans exceed the maximum expe te.ve te:

flow rate._ so an overpressurization 'within-the sheet metal ductwork is not.

These fans can.also be lined up to take c

expected.

4 sup:11es in the event of -a. loss of normal power. power from emergency power-If the radiation levels i

l exceed 40,000 R/hr, the, vent flow will be diverted to the Reactor Building and_

filtered through the Standby Gas Treatment system. This method allows-partic-ulate' plate out in the' Reactor. Building and, in conjunction with filtering,

' reduces the potential release to the stack and subsequently to the environ-ment.

In a design basis accident the PCPL will not be exceeded.

The PCPL can only

.be exceeded in an accident which has progressed beyond the design basis.

Linstiation of venting to maintain the containment pressure below -the PCPL 1,

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'U.S. Nuclear Regulatory Comission 9 '

A08591/Page 7 1,.

October 30, 1989 assures < operability of the vent valves, safety / relief valves, and - structural integrity. of the containment.

Positive effects of venting may include reducing core melt likelihood, reduc-ing the consequences of severe accidents, and avoiding containment failure.

-)'

-However, =the benefits of a hardened vent are limited to-enhancing current capabilities regarding vent paths, which were described above, not provicing a.

)

previously nonexistent i:apability.

During thk-Cycle 12. refueling outage, environmentally qualified, AC powered valve operators were installed on the containment spray MOVs.

The benefit'ofi qualifying these valves was the ability to expand the ' use of drywell. spray under the revised E0Ps.

Although not a benefit to the -TW sequence, drywell spray can have a significant impact on the plant response to a LOCA, Use of~

spray can be very effective in reducing containment pressure and temperature and in scruebing the containment atmosphere.

NNECO's Artroach to Hardened Vent Path Issue NNECO telieves the risk of the TW sequence is low.

Initially,. we also.

believeo that a hardenec vent would further reduce the risk of this sequence and provice a small improvement in overall plant safety at Millstone Unit No. 1.

However. uncertainty over the purpose of and design. criteria for' the hardened vent, plus consideration of previously unidentified technical issues related to interaction of the vent with existing plant systems, have 'shown.

l that additional. analysis is - required.

It is not unreasonable to postulate' that tne ' systems interaction effects coul.d have a potential negative safety, inca:t.if not designed properly.

This additional analysis will focus-on-l determining whether the installation of a hardened' vent will reduce the CMF at illstone Unit No.1. for the TW sequence.

At this time, it -is not 'readily' W

i

'accarents that 4 hardened vent will appreciably reduce the CMF,.'especially g uen the existence of the IC at Millstone Unit:No. 1.

1 NNECO intends to Utilize the IPE as the framework for the analysis.

We_have L

committe: te expanding the Millstone-Unit No.1 PRA model -to -meet the prov1-stons cf Generic Letter 88 20.

Following the actual IPE, we intend to evalu.

(

ate any potential design changes through ISAP Topic 1.113, consistent with accepted past practice and the-pending' license amendment.

Our - decision L

- regarding installation of a hardened wetwell vent will be based on-these results.

We are prepared to commit to'the installation of a-hardened vent, t

provided a functional and cost effective design is developed for the reduction:

j of the CHF fer the TW sequence.

Although our decision to install a hardened.

vent will be tesed on its ability to reduce the CMF for the TW sequece, once-that cecision is made, we intend to maximize the benefit of a hardened vent to l

further reduce the CMF from other accident sequences and to protect contain-J ment integrity.

We'believe this approach to be consistent with the. philosophy behind accident management.

Notwithstanding the above, NNECO will be working with the 8WR Owners Group to.

develop general design criteria and will tailor these to the ' Millstone Unit i,

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' U.S. Nuclear Regulatory _Comission

/,% 6 A08591/Page 8'

{

October 30, 1989 s

No. 1 design specifics. _ lt is anticipated--that such design criteria will be

-f available for-NRC review by April: 30, 1990.

Presuming we decide to volun-(tarily'co m t to the installation of the hardened vent, plant specific design details will: be developed as we-complete the appropriate portions of the IPE~

ar.d' study the possibility of systems interaction effects between the vent and the existing' plant design.

Consideratior u ll also be is able to operate.and reclose,given to the conditions under which the vent as well as potential failure modes of the vent; e.g., random, seal-degradation, and steam.

If the design criteria and related issues can be successfully resolved, and we concludef the hardened vent could be beneficial to the operators, we will-

'l senecule-installation in accordance with ISAP.

Our intent would be to com-i plete installation during the second refueling outage from the date of this letter.

This is consistent with the Commission's goal of. having this issue resol ~ ed within capproximately three years.

NNECO is determined to satisfac-

~

v torily address this issue and will inform the Staff of our decision not later than December 15, 1990. We therefore recomend that the Staff not undertake a backfit analysis for. Millstone Unit No. 1-at this time.

In case the above approach.is not acceptable to the Staff, we are including our initial cost. estimates in Attachment 1, as requested in Generic Let.

ter 89 16.

These values 'are not considered bounding as the esticates are' based sole'.y on concepts, rather than firm design criteria.

As V,e Staff is 1

aware, costs can = increase significantly as the design progrer ses from tre conceptual phase to the plant specific final design.

NNECO strongly prefe s i

to make any appropriate modifications only once. so the risk / benefit of L

alternative cesigns wi,11 be fully considered before any implementation plans l

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are mace.

I Su- !"v and ' CcMiusien

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1 Oar centinuing objective -is-.to expend our resources where' the everall: safety return at the~four units operated by Northeast Utilities is greatest. WeLare nct yet cen,ince0, however, that installation of a hardened vent'at-Millstone

=1 Unt S:. I weald be in furtherance of that objective, nor do we believe that i

installation of a hardened vent is neededi to nhieve " adequate protection,0 especially given the existence'of.an IC at Millstone Unit No. 1.

As a multi-r, p

. unit utility, our objective is to-thoroughly evaluate the safety issues at

'each plant prior to expending our limited resources.

Towards that enc, imeroving the calculated CMF at the Haddam Neck Plant, estimated ~to be above.

10 '/ year 15 a current high priority corporate objective.

l.

l' At Millstone Unit No.1, NNECO has utilized the ISAP-process and PRA analysis to maximize returns, in terms of plant safety and performance. Our belief is that the 'overall safety status of the various factors minimizing.the impor-f, tance of hardened vent capability is very positive.

Several factors contrib Uting to.this positive status are:

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- 'O,$UNuclear Regulatory Connission S'

A08591/Page 9 October 30, 1989-o

. Low Cale'ulated CMF l

o Low ~$80 Contribution to'CHF implementation of Revision,4 of EPGs o

Existing Venting Capability' 11 o-o-

-Isolation Condenser V

o' Drywell Spray Capability y

Significant AC Power Options 4

o u

Cross Tie to.two Millstone-Unit No. 2 Diesel Generators s

i 23-kV Flanders Line Feeding Emergency Station Services Trans-E former

+

NNECO's' positive experiences with liAP have demonstrated the program to b i

resource efficient ' and cost ef fective process for implementing plant; improve-

~

ments, while the Millstone Unit No.1 PSS has been shown-to be an efftctive

-yarcstick for helping to determine the priority of each action.

Expansion'of the. PRA moce),.in conjunction with:the IPE, is expected to provideiquantita--

ti'e ~asults v

on _which a more rigorous evaluation of-recommanded Mark I improuments,- including hardened _wetwell vent,. can be based.

actual IPE, NNECO will evaluate any potential ~ design changes through ISAP.

We':believe that the continued hse of the proven..ISAP program will allow y

Millstoiie Unit No. I to maintain its record of operating safely, in accordance with the' NortheastE Utilities' wnere tne greatest benefits can be' achieved. philosophy of ~ spending' its finite reso s

i!f,you sncaid have any questions, please contact my=staffi h.

Very truly yours, 4

NORTHEAST-NUCl. EAR ENERGY COMPANY s

4/

E. EjWezka f/

fs Senior Vice President i

4 cc:

W. T. Russell, Region I Administrator l-E M. L. Boyle NRC Project Manager, Millstone Unit No. 1 W. J. Raymond, Senior Resident-Ir.spector, Millstone Unit-Nos. 1, 2,,and_3

. STATE 0F CONNECTICUT) n

) ss. Berlin COUhT1 0F HARTFORD )

h Then personally appeared before me, E. J. Mroczka, who being duly sworn, did state that he is Senior Vice President of (2rtheast Nuclear Energy Company, a Licensee herein, that he is authorizeo to execute and file 1the - foregoing n

b information in tha name = and on behalf of the Licensee herein, and. that the statements contained " said information are true and correct to the best of l-nis knowledge and belief.

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Millstone Nuclear Power Station., Unit No. 1 N.1 Response to Generic Letter 89 16 I

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_ Installation of a Hardened Wetwell Vent CostEstimades E

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w October 30, 1989 Millstone. Nuclear ^ Power Station Unit No. 1 s

m Response to Generic Letter 89 16 L

Installation of a Hardened Wetwell Vent Cost Estimates A.

Cost Estimate for Hardened Vent Path The--initial' cost ' estimate for installation of a hardened vent-path. -

similar to the' design utilized by Boston Edison, is $1.1 million.

Thei conceptual design uses existing AC powered containment isolation 2 valves to the maximum extent possible.

As mentioned in the cover letter, additional-engineering. is needEd to establish approcriate design criteria to ensure that the hardened vent system fully meets its intended function.

The cost estimates to install a properly engineered hardened vent at Millstone Unit No; I may. increase

~L dramatically based on the. specific design requirements for -flow, pres-sure, single f ailure, environmental qualification for temperature and1 i

radiation, and seismic support. Also, the cost estimates-do not consider

~

potential.. backflow damper work, nor the implications to. Millstone Unit Nos. 2 and 3, located at the~same site.

B.

Incre emtal Cost Estimate for AC Indeoendent Desian The incremental' cost : estimate for an AC independent designEis an: acci-1 tional 50.6 million, for a total of $1.7 million.

The. conceptual! design 71 incluces installation of several DC powered motor operated valves, but

-does not consider the issue.. of 24-hour -post SB0 availability.

gO_,

SECY 89 017 states that licensees imple.wnting' the SB0 rule' by..use of_ an AAC source need-1not provide additional power ' supplies., provided, the chDacity of.the AAC is sufficient for the requirements of both-the SB0 i

l rule and the' vent design.

Although no analysis has yet,been performec, Y

it is expected that the above criterion would be met.

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5000-69-1832 v.3..N.:; ear Pe; alas:ry Cer.-;ss :n A :nt- 00:; rent Cent :1 Oesk

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Oent.enen:

-t S;::e:::. Cyster Creek Nuclear. Generating Station L

ket No.'50-219 Pesp:nse :: Generi Letter 89-16 3

Marx : C:nta nren Hardened Vent

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In.s.'e::er ::ns ::utes CPU Na:; ear's resp:nse :: Gener.: 1.e::er ii-;f fer-:..e Oyster Oree< Na:; ear-Genera:In; Sta ':n.

The genera: letter'was re:e:ved :y

,7'..S.:. ear :n Septer:er ;4, 1959.

We ::r.:: :: Install-a nardened vent prev;ded the resalts of gr ;PI snr'.

na:

..s a. appr:priate ::sk redu::.:n measure.

We w 11 netafy y:a cf 0.:

n:'.st:n as scen as avat;acle : :

later enan cur :PE scheda;e.

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Wet e.. ven: n; :apac u lty currently,ex;s s at; oyster Creek.

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(;;a:e a.d Pro:edures are e.-

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perat:rs are trained :: ven: ::ntain ent ;f needed :: 4:ntain-

-::rta:- en 4ntegrat,y.

The vent patn as n:. " hardened" and snus : uld ce expe::e :: re'.teve pressure : :he rea:::: outicang af used at n:qn pressare.

5 M.ever,

..e a::11:y t: ree. eve de:ay nea: :nreugn it:s path andv:r av::.:

3;r..ary ::ntaar. en pressure-lamat

..e is provided my tne current-des.,...

4

Eased

n tr.e NRC Staf f's analysis an S!';Y-89-017 and en discusstens to: ween;3W2

J. ors'

rtup representatives and the Staff, we understand the princi N

pal,rask i

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re:.::::n tenef t; associated with venting is expected for plants whose

2.:a;a:ed :=re-damage'fregaency (COF) is.demanated by acciden sequences l

W

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.g.* ng Term Less of Decay Heat Removal (TW).

Oyster Creek is

!!i s.:s:a ::4.;y'dtfferent;frem tne planta represented n.the Staff's reference W

a.alysisi

has severa; destgn features snat would'autstant

4;;y re: ace :t.o-f

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l' Jade;;.::d :f tw seqdences.and thus reddee the menet::

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cf the hardene von:.

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S pe :i'i:a lly, the oyster 1 Creek plant

. System f.:CS ) shown schematically in Attachmentdesign employs an Isolation Condenser 1.

There are two independent; JCS tra:ns (ene'is pictured), each fully redundant, which are capable of

- Opera:;ng ' independent' of AC power. - The ICS is' a passive system with 4

capatie of removing at least 3 percent reactor power without rejecting heat each train nt:.centa;nment.

Further, in:;uding a scurce_via redundant diesel-fire pumps which arediverse sources independent of tne emergency.AC power cystem.

heat dare :;y_to atmospnere, containment would notSince the isolation condensers rem p"

.s.tk during reacter isolation sequences as long as IC's are available,be chosen as a prim very un;ike;y sequences-invclving loss of both ICs, Oyster Creek-has diverse Even :n' 3

nea: reroval paths not available at plants considered in the NRC's reference analys s.

These include: 1)-four contatnment spray heat'exchangers with serv;:e! water supplies which are powered from diesel generators and are

n=ependent ef-IC's and other paths, and 2) three shutdown cooling loops w::n t

serv; eLWater independent of ICs and containment sprays which could be used-f:r seguences inv:lving' low reacter pressure.

Thus, the likelihood of cces.dara;e-

':r ::n aineen: ever-pressure due to.de:ay heat is extremely small.

n add;;;:n :: these existing heat remeva.' paths, our planned modif :at;en ::

add an a;;ernate AC (AAC) power source during-the 14R outage will irpreve the re;; ate.;;y of AC pcwer :n further assure the availability of these heat rer: val patas in :ne event diesel generaters are unavailable.

Early vintage PP.A-type studies for Oyster Creek did net show decay heat removal sequences t:

e-s;;nifi: ant centributors.

It is cur expectation'that the IPE, wh h is

.rrently underway, will also snow that fra:: :n.cf ne'tetal calculated CCT.

TR sequences will represent a very ;:w Therefore, the benefit of a hardened

enta.nrent vent h

is substant; ally less tnan plants w hout ICS, lLj;

: p a r a.'.' e ' w; n the.:PE effort to eva;; ate :CS use, we will be werk n; w :n n* SWE 0.ners ' Grcup to devele; gener:e design criteria for :ne aar:ened vent.

Tne owners' Group ant;:1 pates that gener; har:ened vent des;;n crater.a

--;.. :e. ava;' a:;e f or NRC reviewicy Apr;l ~30, 1990.

We nave perf:rmed preliminary engineering to develop a concept'fer a hard ven:

wn;:n s s:neratically shown in Attacheen 2.

ened This design is essentially i

sne sare as :nat wn ch was described in Enclosure l~to the' generic letter, except f:r the use of remotely operated valvas:instead of rupture;dises, whien

-e currently consider, to be more technically appropriate if a hardensd vent q tt

nsta;;ed.

as L

Our preliminary cost estimate f or implementing this design is $2.;

L Fr;;.';:n cased on installation in 1993.

Our preliminary cost estimate for

[

m:: fy:ng the design to borindependent.cf!AC power is an incremental increase j

'! $0.5 eClion.

The treakdown of these estimates by major element b

of cost is P

7:a;ned.;n Attaencent 3.

This ces: :n well in excess of that assumed in :ne P

ener;

;etter.

Ma;or etus components include the^ distance and : route of tne ven: pata, :no reactor. building penetration, length of cable / conduit

-the ::ntre' runs ::

cm and'the engineering cf radiata n monitoring and H

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r: sti:n ::ntr:1: features.

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s.:stantially mere shan,the estimate in the generie letter and the j

expe:ted :e..efits would be substantially lower than the reference plant (s) dae ter:gr uniq.e des.gn features.

• hus the cest/ benefit ratto is far different.

(

We w;11-thercug..ly examine the hardened vent issue in the-IPI in the context cf TW an: all ciner sequences. We will also take into account other Mark :

ntaanment.. issues-adentified by the staff.

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October 30, 1989 s

L U.S. Nuclear Regulatory Commission ATTN:

Document Control Desk Nashington, D.C.

20555 i

Subject:

Dresden Station Units 2 and 3 Response to-Generic Letter 89-16 NRC Docket Nos. 50-237/249

Reference:

Generic Letter 89-16, Installation of a Hardened Netwell.Vcnt, dated September 1, 1989.

?

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Dear Sir:

The referenced Generic 1.etter informed the licensees of Bolling Water Reactor (BWR) plants with Mark I containments of the NRC program for

+

disposition of-the-issues related to the Mark I Containment Performance l

Improvement Program'.

The Generic Letter encouraged licensees to voluntarily

' install a hardened. vent under the provision of 10-CFR.50.59,'and requested that, licensees: provide notification of their plans' for this Issue. This letter provides Commonwealth Edison's (Ecison's) response to that request.

SECY 89-017 indicates that the primary benefit of'the hardened vent is the reduction of risk associated with the TW (loss of decay heat removal)-

seauence.

Edison believes that actions necessary to mitigate this sequence:

may be accomplished by taking full advantage of the tsolation condensers already installed at Dresden-Station Units'2 and 3.

Edison is evaluating the use of the isolation condensers as a viable alternative to a hardened vent for the mitigation of a TW sequence. The BWR

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Owners' Group (BHROG).is currently developing generic design criteria for a haroened vent.

It is anticipated that these criteria will'be available-by.-

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< April ~30. 1990, for NRC review.

Edison is participating in the BWROG's

'l efforts in order to provide the rationale for. use of the isolation condensers at Dresden. Station.

This rationale will be provided to the NRC by Edison 60 days following issuance of the BWROG's generic design criteria.

Edison will be performing an Individual Plant Examination (!PE) for Dresden Station to fulfill the reautrements of Generic Letter 88-20, with an expected completion.date of April 19?2. Upon completion of this IPE, a review

,of.'its results may identify additional considerations specific to Dresden Station with' respect to the TH secuence issue, c

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October 30, 1989 l

Edison has comoleted a conceptual cost estimate for a. hardened vent similar-in design to that depleted in the Generic Letter.-'The' cost for this -

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-type of vent design at Dresden Station Units 2 and 3 is estimated at soproximately 52 million.

cost estimate by an additional $1-million.Providing AC power independence could in

?

Edison believes that utilization of the presently installed Isolation i

condensers at Drescen Station significantly contributes to closing the NRC's Mark 1 Containment Per. form'.nce Improvement Program.

s Please dire:t any questions that you may have concerning this response to this office Respectfully, s

3 hf.lY,e,*

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M.H. Richter Generic Issues Administrator 4

' lcc: 'A.Bi. Davis - Regional Administrator, Region III Resident Inspector - Dresden i

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D590-01].

DR}S NUCLEAR REGULATORY COMMISSION I

Installation and Operation of Hardened L

Vent From Suppression pool Airspaces of BoilingWaterReactions(BWRs)with Mark 1 Containments AGENCY:

-Nuclear. Regulatory Commission ACT10ti:

Draft Generic Environmental Assessment and Finding of No Significant Impact.-

f

SUMMARY

The installation of the hard pipe' vent in flark I plants will reduce the environmental consequences of a severe accident involving loss of 1

- long-term decay heat renoval capability and provide a significant. improvement

.in sefety. Insta11ation'or ese of the hard pipe vent will not have any-significant environmental impact'.

j i

The incremental occupational radiation dose for the proposed operation of.

the herd pipe. vent path is insignificant (unmeasurable) because the vent path would'be operated from the control room. The licensees should be able to keep i

the small radietion doses associated with the installation of the.hard pipe vent path within the limits of 10 CFR Part 20, and as low as is reasonably j

achievabic.--Furthermore, the non-radiological impacts of the hard pipe vent path wil1 be insig'nificant. None of the alternatives is practical or reasonable, and_three of the alternatives would produce greater environmental j

y.,

impact than the preposed action.

Addition of the external filter would have n

m the'same environmental impact as the proposed action, but at.an unreasonable

' cot.t for minimal increase in berefit.

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Alternative Use of Resources This action does not involve the use of significant resources beyond the existing resources used for piping and replacement parts at all nuclear plants.

Agencies and Persons Consulted-e

- i The NRC staff is initiating this action based on research performed by the j

Office of Nuclear Regulatory Research.

No other agencies or persons were censulted.

1 i

DATE:

The ccrrent period expires [ Insert a date allowing 60 days for public coment].. Coments recei,ved af ter this date will be considered if it

-is practicci to do so, but assurance of consideration can not be given'except I

to those coments received on or before thir date.

4 ADDFESS:

Send comments to the Regulatory Publication Branch, Division of i

Freeden of-Information and Publication Services, Office of Administration, l

-U.S. Nuclear Regulatory Commission, Washington, DC 20555. Copies of the

-coments may be inspected and copied for a fee at the NRC Public Document Room,-

the Gelman Building, 2120 1. Street, N.W., Washington, DC.

V'

_ j FOR FURTHER INFORMATION CONTACT:

Mohan C. Thadani, Division of Reactor Projects 1/11, Telephone (301)492-1419, or John A. Kudrick, Division of Systems Technology, Telephone (301)492-0871, Office of Nuclear Reactor Regulation, 11.S. Nuclear Regulatory Comissior., Washington, DC 20555.

6 2

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[7590-01]

SUPPLEMENTARY INFORMATION Backoround in SECY-87-297 (Reference 1), the NRC staff presented to the Comitsion its plan to evaluate generic severe accident vulnerabilities of containments..

l The staff's plan included a program for Containment Performance Improvement (CP1). This program was initiated to determine whether there may be generic

- [

severe accident che11enges to light water reactor (LWR) containments that i

should be assessed to ascertain whether additional regulatory guidance or I

requirenents concerning containnent features are warranted. The staff concluded that such assessments are needed because of the relatively large uncertainty in the ability of some LWR containments (that is, Mark 1) to successfully survive some possibic severe accident cha'11enges (as indicated by draft HUREG-1150 (Reference 2)). The CPI progran is intended to resolve hardware and procedural issues related to generic containment challenges. The staff presented its findings related to the Mark 1 CPI program to the Comissien in SECY-89-017 (Reference 3),detedJanuary 23, 1989.

1r cne of the findings, the staff concluded that properly implemented venting can significantly mitigate potential accident risks.

The capability to vent has long been recognized as important in reducing-risk at boiling water reactor (BWF.) Mark I f acilities for accidents involving loss of-ability to remove long-term decay heat (TW). Controlled venting at pressures close to the containment's pressure limits can prevent the long-term overpressurization and failure of containment, the failure of emergency core F

cooling system (ECCS) pumps from inadequate net-positive suction head, and the failure of the automatic depressurization system (AM) caused by the failurt cf 3

l

[7590-01]

ADS valves to operate. Venting of the containment is permitted in the BWR emergency operating procedures, A. vent path from wetwells of the containments exists for some Mark 1 facilities. However, this vent path includes a ductwork system that has a low design pressure of only a few psi. Venting under high-pressure conditions (as would be required for accidents involving high-pressure challenges, either before or after core melt) could fail this ductwork, releasing thecontainmentatmosphereintothereactorbuilding(witheventualreleaseto the environment), and potentially contaminating or damaging equipment needed for eccident recovery.

In addition, with the existing hardware and procedures at some plants, it may not be possible to' open or to close the vent valves during-ce'rtain accident sequences. The inability to operate the vent path valves could result in uncontrolled release of containment atmosphere to the reactor building through the failed sheetmetal c'uctwork. Therefore, venting through-a sheet retal ductwork path, as inplemented at some Mark I plants, is likely to greatly hamper or complicete post-accident recovery activities, and is viewed by the NRC staff as inadequate for minimizing the risks to the public health and safety.

For high-pressure venting to be effective, the entire vent system must be strengthened to withstand the expected venting pressure. On July 11, 1989 (Reference 4), the Commission endorset, the staff's view that the Mark I design should include a hardened vent from the airspace of the containment wetwell, j

and directed the staff to require a hardened vent capability for all Mark I plants for which the requisite modif'ications could be shown to be cost-effective.

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[7590-01]

Description of the' Proposed Action:

Installation of a liardened Vent The NRC staff's safety evaluation report (Reference 5) approved Revision 4 oftheEmergencyProcedureGuidelines(EPGs)thatincludedthestaff'sapproval for venting BWR Mark I containments. This approval indicated that venting with the existing systems could reduce the likelihood of core melt and, in extremely I

rare cases, could help avoid uncontrolled releases of radioactive materials during severe core damese accidents.

Since the issuance of Revision 4 of the FPGs, additional insights indicate that a venting strategy that has a potential to breach the vent path inside the reactor building could have significant detrimental effects on (1) radiation exposure impact on personnel, (2) potential plant recovery actions, and (3) public risk. A hard pipe vent capable of withstanding the anticipated severe accident pressure loadings would eliminate

'these disadvantages of using a vent path containing sheetmetal ductwork; The use of the containment vent to prevent a core melt accident, by reducing containment pressure, would result in the release of very low levels of radioactivity associated with the reactor coolant. The; reactor coolant steam would be released to the suppression pool that would retain most of the a

fission products.

In the unlikely event of a core melt accident, venting of the wetwell eirspace would provide a scrubbed venting path to significantly reduce the release of particulate and volatile fission products (radioactive materials)totheenvironment. Only the noble gases would escape to the environment without any attenuation. Venting would reduce the likelihood of a late overpressure failure of the containment and would reduce offrite consequences

'for severe accidents provided that the containment shell does not fail.

5

[7590-01]

Iftheshellfailsbecauseofacoredebrisattack(shellmeltthroughby core melt releesed to the containment floor); venting will provide little benefit because fission products would be released directly into the reactor building. However, if shell failure was delayed for a period of a few hours (for example, by the addition of containment spray water over the molten core debris released to the containment floor), significant scrubbing of radioactive material would still take place. A recent analysis has been performed on the effects of water on core debris in the drywell (Draft HUREG/CR-5423, "The Probebility of Liner Failure in a Mark 1 Conteinment" currently released for peer review). Preliminary results indicate that the presence of substantial quantities of water in the containment floor area on top of any molten core debris (two to threE feet of Water overlying the C0re debris), that Would result from injection of water from available sources, will very likely prevent containment shell melt through and failure as a result of a core debris attack.

The. overlying pool of water will also provide scrubbing of fission products released in the aerosols generated by core melt and concrete interactions.

As-proposed in SECY-87--297 (Reference 1),.the installation of a hard pipe

.to bypass the ductwork from the wetwell airspace-_to the plant stack could include (1) additicnal isolation valves tc isolate the ductwork path from the har'd pipe vent path, and (2) radiation monitor (s) to monitor any offsite releases of radioactive materials, in case of venting. The proposed action i

would prevent failure of the vent path inside the reactor building and, in the unlikely event of core melt, would result in release only of residual fission products (not scrubbed by the suppression pool) through the stack. Because the vent path is not expected to fail inside the reactor building, personnel would be able to repair equipment and perform other plant-recevery activities,

[7590-01)

-provided the levels of radiation in the containment are not excessive. Further-more, because the environmente.1 conditions in the reactor building would not be harsh, important equipment would not be expected to degrade or fail.

In the proposed action, ell potential releases through the hard pipe vent will'be scrubbed by the suppression pool water that will reduce the radioactive material released to the environment, but will not decrease the release of the i

noble gases. The effectiveness of the scrubbing is affected by the temperature of the suppression pcol water. Depending upon the temperature, the decontamination factors could vary from three orders of magnitude to one order of magnitude, bet over the course of the eccident, the effective decontamination factor would-be about tro orders of megnitude. However, as long as water is present, all

]

releases to the vent will first pass throuuh the water that will retain substantial fractions of radioactive r.aterial. Additionally, the'use of the hard pipe vent could prevent or delay core degradation for those accidents where containment failure retuits in core degradation, as previously explained.

The estimated reductions in the values of the total core damage frequency per reactor-yer.r are shown in Teble 1 for each Mark I plant. The risk reduction in man-rems'per reactor year is also shown in Table 1 (The beses and assumptions for the staff analyses are presented in Reference 6). The hard vent path would also provide additional risk reduction for those accidents where core melt has occurred, because the st,ppression pool would scrub the radioactive material released by molten core.

The PRC staff estimated the costs for installation of the hard pipe vent path to be about $750,000 (Reference 7). Costs were also provided by the licensees for the Dresdor,, Fitzpatrick, Millstone, and Oyster Creek facilities.

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[7590-01]

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-The costs are minimal when-compared to the operating expenses of the-plants and are not excessive when compared to'the significant enhancement of safety-achieved by the proposed action. The NRC and the licensees' costs are summarized in Table 1.

Environmental impacts of Installation and Operation of a Hard Vent j

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Radiological Impac:

t The radiological impacts of installation of a tiard pipe vent system should--

not be significantly different from other operational modifications that occur at facilities such as reactors with Mark I containments. For example, a conceptual analysis of radiation exposures for inste11ation of a filtered vent at the Limerick Generating Station indicates that annual radiation exposures

-(assuming 20 years of remaining plant life) would not exceed 2 man-rems per reactor-year._ The small radiation dose associated with this proposed plant modification will not affect the licensee's ability to maintain individual

? occupational doses within the limits of 10 CFR Part 20, and is' expected to meet' the'criteriafortherequirementsofaslowasis_reasonablyachievable(ALARA).

Each plant contains radioactive waste treatment systems that are designed L

' to collect and process the gaseous, liquid, and solid. waste that might contain radioahtivematerial. The proposed installation of a hard pipe vent will not-affect any waste treatment systems or their effluents under normal plant conditions or under design basis accident conditions.

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[7590-01]

Installation of the hard pipe vent path should not significantly increase the radiation dose to operating personnel or the public. Any increased doses associated with the testing of the additional isolation valves should be minimal and, in most cares, insignificant.

Thus, we have concluded that the proposed installation of the hard pipe vent will not result in-any significant increase in radiological impacts to workers or the public, Because the operation of the wetwell vent system is postulated for extremely rare severe accidents, the impacts of the use of the wetwell vent system are discussed in terms of environmental risks.

As stated previously, the venting from the wetwell airspace is intended to (1) reduce the risk of over-pressure failure of the containment and subsequent damepe to the reactor core, end (2) provide a scrubbed pathway (to decontaminate effluents) for conteinment pressure relief for rare situations involv'ing core damage. Table 1 shows a listing of-reduction of potential risks for all Mark 1 facilities caused by venting prior

'to core damagt.

The reductions in potc;;tial risk are calculated to range from 15.3 to 281.9 man-rems per reactor year.

For rare situations where core damage.

could occur, venting could prevent containment failure and unmitigated release of fission products to the environment.

Venting through the suppression pool will ensure that most of the radioactive materials, excluding noble gases, would be trapped in the suppression pool and would not be released to the environment. Therefore, the use of the vent system would reduce the radiological risks posed by severe accidents involving core damage. These additional-benefits of verting have not been included in Table 1 results.

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[7590-01]

L 1,3 Based on the preceding discussion, we conclude that there will be no incremental environmental risks posed by operation of the wetwell vent system.

Non-radiological Impacts' The non-radiological impacts of installation of hard pipe vent system are not expected to be different from other operational modifications that occur at facilities such as reactors with Mark I containments during routine plant outages.

No non-radiological effluents are expected to be affected by the installation or use of the hard pipe vent. The proposed plant modification and use of hard pipe vent will not require any change to the national' pollution discharge elimination system (NPDES) permit.

Therefore, the staff conclu' des that the non-radiological environmental impacts of installing a hard pipe vent will be insignificant.

Alternatives Considered b

.I To prevent or delay containment failure caused by overpressurization, the NRC staff censidered the following alternatives to the proposed action:

1.

The containment pressure could bc relieved using the existing ductwork.

vent path (the'"No Action" option).

2.

A hard pipe path to an external filter could be installed.

L 3.

An alternate means of removing the decay heat either from the reactor L

-or from containment could be installed.

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[7590-01]

4 3 4

Venting of containnent could be prohibited.

Each of these alternatives to the. proposed action is discussed in the a

following paragraphs.

L, ExistingDuctwork.VentPath(NoActionOption)

.i This alternative consists of no action and continued venting of the containment through the existing ductwork. However, the existing ductworks are I

designed to withstand a pressure of a few psid (Reference 8). The venting.

pressures expected during some accidents will be substantially higher than the ductwork design pressure. Consequently, venting could result in failure of the ductwork and a' direct.releese of reactor coolant steam into the reactor building.

i The discharge of this high-temperature steam and other gases over an extended period of tire could pose a thret.t to the availability or performance of.

safety-related equipment.

In the event of core melt, the threat would be even~

l greater, because substantially large amounts of radioactive materiait will be released with the steam to the reactor building. Electrical cables, motor-operators on valves, and relays could fail under these environmental conditions.

Adverse environmental conditions would also complicate personnel entry into the reactor building. Calculations from a study that examined venting during an anticipated transient without scram (ATWS) sequence indicated that e bevere 1

environment (high temperature and radiation _) would be present in the reactor building'duringventing(Reference 9). The discharge of hydrogen under core I

~

melt conditions could result in hydrogen burns or detonations inside the reactor building. This environment could hamper recovery efforts by preventing

'1 11 l

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[7590-01) q,

. personnel access to the reactor building and preventing: repair of systems needed to terminate the accident. For these reasons, the existing Mark 1:

designs do not ensure an adequate reactor building environment after a severe 1

accident to permit personnel entry to regain control of the facility and do not maximize the potential reduction in environmental risk. Thus, the staff-has concluded that the no-action alternative is unacceptable.

i installation of Hard Pipe Vent to External Filter System This alternative is the same as the proposed action with addition of an external filter.

However, the external filter would not significantly increase i

removal of radioactive material because the suppression pool would remove

-i nearly all naterial that could be removed by filtration. Consequently, the

' additional reduction in risk caused by an external filter system is expected to

.be small, tioreover, an external filter would not yield an incremental reduction

?

in the. core. damage frequency beyond the reduction obtained with the hard pipe vent'alene.

In both cases, there would be no retention of noble gases.

External filters have been estimated to cost $20 million (1982 dollars)

L (Reference 10) to $65 million (1987 dollars)-(Reference 11) for the Filtra design. Because the incremental benefit is very small_ compared to the proposed f

l

action and the iiicremental cost is very high, this alternative is not considered practical or reasonable, i

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Installation of Other Means of Decay Heat Removal In lieu of venting containment, an additional decay heat removal system could be provided to remove the heat from either the reactor or the containment, or a system that has not been previously accounted for could be used on an ad hoe basis, such' as the reactor' water cleanup system. Installation of a new system was considered in NUREG-1289 (Reference 12), which is associated with Unresolved Safety Issue A-45, " shutdown heat removal requirements." The instal-lation of a new decay heat removal system was not found to be cost beneficial in NUREG-1289. The use of-another, previously unaccounted-for system was estimated to rcouire unusual or unplanned system piping line-ups, which, if performed incorrectly or inappropriately, could reduce the likelihood of j

l

-accident recovery with normal systems or create a new and unanalyzed accident sequence (Reference 13). Therefo'ro, this alternative is not considered practical j

or reasonable, j

No Venting of Containment This alternetive would remove the guidance in Revision 4 of the Emergency

' Procedure Guidelines (EPGs) that instructs the operator-to vent the containment under certain conditions.

In the event of the loss of long-tern decay heat -

1 removal capability without drywell f ailure, the containment drywell will probably fail-because of overpressurization. The drywell failure could have a significart effect on the ability to return the plant to a tafe and controlled cer.dition and would result in an increase in risk to plant personnel and to the 13

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[7590-01) public(Reference'14). Therefore, this alternative is not considered practical or reasonable.

Findino of No Significant Impact The staff reviewed the plant-specific features in conjunction with the proposed hard pipe vent path modification relative.to the requirements set-forth in 10 CFR Part'51.. From the environmental assessment, the staff concluded that there are no significant radiological or non-radiological impacts associated v:ith the proposed action end that the proposed modification will not have significant adverse effects on the quality of the human environment. Therefore, the Commission has determined, pursuant to 10 CFR 51.31, not to prepare an environmental impact statement for the proposed plant modifications.

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[7590-01)

Table 1 1

Potential Installation TW Frequency Risk Reduction Costs Plant Name (perreactor-yeer)**

(man-rems /ry*)**(million)**

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Browns Ferry 1 2.3 E-05 32.7 0.75 I

Browns Ferry 2 2.3 E-05 32.7 0.75 Browns Ferry 3

?.3 E-05 32.7 0.75 Brunswick 1 4.5 E-05 44.0 0.75 Brunswick 2 4.5 E-05 44,0 0.75 Cooper 4.5 E-05 45.6 0.75 Dresden P' 1.4 E-05 50.2 1.00 L

Dresden 3 1.4 E-05 50.?

1.00 l

Duane Arnold 4.5 F 55.0 0.75 Fermi ?

4.5=E-05 192.4

-0.75-Fitzpatrick 4.5 E-05 65.5 0.68 L

Hatch 1 4.5 E-05 39.2 0.75 Hatch ?

4.5 E-05 39.2 0.75 Hope Creek 6.3 E-05 281.9 0.75 Millstone 1 1.4 E-05 35.1 1.10

- Monticello 4.5 E-05 33.9 0.75:

Nine Mile Point 1 1.4 E-05 15.3 0.75 Oyster Creek 1.4 E-05 55.4 1.50 Peech Bottom 2 3.6 E-06 15.5 0.75 Peach Bottom 3 346 E-06 15.5 0.75 L

Pilgrim -

2.3'E 31.2 0.75~

Quad Citics 1 4.5 E-05 94.1 0.75 Quad Cities 2.

4.5 E-05 94.1 0.75 Vermont Yankee 2.3 E-05 28.9

. 0.75-l t

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reactor year l

-'** Reference 15 l

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[7590-01]

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REFERENCES l

1. - -SECY-87-297, U.S. NRC, " Mark I Containment Performance Program Plan,"

Y. Stello to NRC Commissioners, December 8, 1987.

L l

2.-

NUREG-1150, U.S. HRC, " Severe Accident Risks: An Assessment for Five U.S.

Nuclear Power Plants " June 1989.

1' 3.

SECY-89-017, U.S. NRC, " Mark 1 Containment Performence Improvement Program,"

't V. Stello to NRC Commissioners, January 2',1989.

a 4

Memorandum from S. J. Chilk to V Stello, "SECY-89-017 - Mark'1 Containment

-Performance Improvement Program," July 11, 1989.

j 5.

Letter from A. C. Thadani to D. Grace, Chairman, BWROG, " Safety Evaluation of BWR Owner's Grcup - Emergency Procedure Cuidelines, Revision 4, NEDC-31331, March 1987," September 12, 1988.

E 6.

~ Memorandum from M. Cunningham to W. D. Beckner, " Reduction in Risk from the Addition of Hardened Vents in BWR Mark I Reactors," October 19, 1989.

7.

Menorandum J. G. Partlow to T. E. Murley, " Licensees' Responses to Generic Letter 89-16 Related to Installation of Hardened Vent," November 9, 1989.

9 8.' -

NUREG/CR-5225, U.S NRC, "An Overview of BWR Mark 1 Containment Venting Risk Implications, November 1988.

16

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[7590-01)

9..

Harring, R.M., " Containment Venting as a Mitigation Technique for BWR Mark 1 Plant ATWS," 1986 Water Reactor Safety Meeting, Gaithersburg, Maryland, October 1986.

10.

K. Johansson, L. Nilsson, and A. Persson, " Design Considerations for implementing a Vent-Tilter System at the PARSEBACK Nuclear Power Plant,"

International Meeting on the Thermal !!uclear Safety, Chicago, Autost 29 -

September 2, 1982.

11. Lorg Island Lighting Company Presentation, "Shoreham Supplemental Containment System," April 19E7.

1

12. IlUREG-1289, U.S. NRC, " Regulatory and Backfit Analysis: Unresolved Safety Issue ' A-45, Shutdown Decay Heat Removal Pequirements," November 1988,

. v

13. Letter from J. Dallman to J. Ridgely, "A Preliminary Assessment of BWR Park 11 Containment Challenges, Failure. Modes, and Potential Improvements in Performance," May 10, 1989.

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14. NUREG/CR-5225, Addendum 1, "An Overview of BWR Mark I Containment Venting j

Risk Implications, an Evaluation of Potential Mark I containnent Improve.

ments," June 1989.

l l

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15. Memorandum from Warren Minners to Ashok C. Thadani, January 8,1990.

Dated at Rockville, Maryland this day of

,1990.

]

For the U.S. Nuclear Regulatory Commission l

Thomas E. Murley, Director y

Office of Nuclear Peactor Regulation l

l l

l 2

NOTE: All referenced documents are available for public inspection and copying for a fee,in the Commission's Public Document Room at 2120 L Street, N.W., Washington, DC 20555.

18

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