Download - Pennsylvania EALs,in j · Pennsylvania Power 8 Light Company Two North Ninth Street ~Allentown, PA 18101-1179 ~ 215/774-5151 Robert G. Byram Senior Vice President-Nuclear 215/774-7502

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DOCKET g0500038705000388

ACCESSION NBR:9401140133 DOC.DATE: 94/01/05 NOTARIZED: NOFACIL:50-387 Susquehanna Steam Electric Station, Unit 1, Pennsylva

50-388 Susquehanna Steam Electric Station, Unit 2i PennsylvaAUTH.NAME AUTHOR AFFILIATION

BYRAM,R.G. Pennsylvania Power 6 Light Co.RECIP.NAME RECIPIENT AFFILIATION

MILLER,C.L. Project Directorate I-2 ~/~ R

D

NOTES:

SUBJECT: Forwards responses to NRC questions, revised pages toemergency plan, revised pages to technical basis 6 proposedEALs,in response to NRC 930609 request for addi info on revto emergency plan which implemented NUMARC NESP-007.

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INTERNAL: AEOD/DOA/IRBNUDOCS-ABSTRACT

EXTERNAL: NRC PDR

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CLARK,R

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Pennsylvania Power 8 Light CompanyTwo North Ninth Street ~Allentown, PA 18101-1179 ~ 215/774-5151

Robert G. ByramSenior Vice President-Nuclear215/774-7502

JAN 05 1994

Director of Nuclear Reactor RegulationAttention: Mr. C: L. Miller, Project DirectorProject Directorate 1-2

Division of Reactor ProjectsU.S. Nuclear Regulatory CommissionWashington,-D;C. 20555

SUSQUEHANNA STEAM ELECTRIC STATIONRESPONSE TO RAI ON REVISION TOEMERGENCY PLAN WHICH IMPLEMENTEDNUMARC NESP-007PLA-4032

Docket Nos. 50-387and 50-388

Dear Mr. Miller:

This letter responds to your StafFs request dated June 9, 1993 for additional information regardingPennsylvania Power &, Light Company's revised Emergency Action Levels in the Emergency Plan.

This letter contains four attachments which are as follows:

Attachment 1 - Responses to the StafFs questionsAttachment 2 - Revised pages to the Emergency PlanAttachment 3 - Revised pages to the Technical BasisAttachment 4 - Proposed Emergency Action Levels

The following are additional revisions to the EALs as a result of PP&L's internal review:

~ EAL 1.1.1.a - The Technical Basis has been revised to reflect the correct parameter.~ EAL 1.2.3.a - The containment high radiation monitor reading was revised from 4000 R/hr to

2800 R/hr to provide a more realistic value.~ EAL 1.2.4.a - The containment high radiation monitor reading was revised from 35,000 R/hr

to 20,000 R/hr to provide a more realistic value.~ EAL 2.1.3.a - The words "cannot be maintained above" have been replaced with the word

"below".~ EAL 2.1.4.a - The words "cannot be maintained above" have been replaced with the word

"below"~ Table 4.1.1 - This table was revised to correct the max safe area temperatures.~ EAL 5.2.3 - This EAL was revised to incorporate the new 10CFR20 methods for dose

determination.~ EAL 5.2.4 - This EAL was revised to incorporate the new 10CFR20 methods for dose

determination-"- I 03" 9

OSO00387PDR ADOCKF

-2- FILE R41-2 PLA-4032Mr. C. L. Miller

Ifyou have any questions, please contact Mr. C.T. Coddington at (215) 774-7915.

Very truly yours,

Attachments

cc: SNRCIlocnmenCControl Desk (original)NRC Region IMr. G. S. Barber, NRC Sr. Resident Inspector - SSESMr. R. J. Clark, NRC Sr. Project Manager - Rockville

nv

n

1.Q REACTOR FUEL

I

)Itf

IO

l.l.l.aChcalttry datcral ass thatReactor Coolaat Actfrltyexcced5 tathattal5peclflcatfoa Ll~ It5

l.l.l.byalld 0/f-Cat PretraataeatHoaltor Hlth Hfth AadfattrmAlera

roca tare1.s.ttl xcl/po

~ct/cc erI/el(bono

teolvoloo\Io41ao.ill

nr 100/(aCISor erm

I 2.1 Hor»

V»um observsaon cc ~ sustnn«Luncontosed water level decrease

ALL ln 0» Spent FOH Pool wlh ashrsrpmad tuel assembesremarxnp covered by »ates

L12

Reactor CIIHant Activity4S DeterounedbySamfre

axanotst

1.2.2

IPS. Vahf Contshunent Hloh~ AadlaUon Monitor

vr/nanny c«aonvn«r tuna«mneo»oodr 21$ 'F. ~ 2$ nrvo«rococoroco«moor I»'«crore»nuory ln me«loom» ro onxor Idxnrucou onoc

I.as1.2+ Veld Containment Hlph4S Radiation Monilor

> 2.800

1224ALL Vntaaianed Vasd nsnr fuel COUCONy

monhor (ARM 12 or Hesah PhysicsSurvey)

IS2 b

ALL I/~vaad spent tuelcritloaaiy monhor (AAMMorHeaah Phyucs Survotf

IS20ALL I/nptsnnad. vasd re/uet Hoor

process red mon|tot (see fatso IJf

130 Itone

> 250mnlhr

> 250auMO

Tatse IS

Refuel Floor ProcessRarpsdon Monhors

Retest Floor Iadh Exhaust DuctAad Monsor

Rasread Access Shah Exha altDuct Aad Mcnaor

Refuel Fhor Was Exhausl DostRad Monacr

I,2+ Reactor Coolant Activity4 Determrnedby Santae

> 1000

axsnoos 1

12SbISA Vtad Contalnmenl Hhfh

~S RadlaUon MonhorAND

Primary Con(alnment InteprUyCannra be nraanr»dvyFrvn«y Conrnrvn«u onv»»mlenreouo 215 F, ~ 15 rurr c«rocton/ocr«nuor te «ar«ra»nuorr oo I»Innrotco ro «root Inoon%on onoc

CD

Ut

IM

U

I.IA

Reactor Coolant AcdvttyDetermmof by Samfce

ANO4S

Primary Containment Inleprltycannot ba Mavcau»d

> 2000OCusm

axrno tat

1.244

IAA Vatd Contalnmenl Hlph4 Red MUon Monitor

1.2.4 b

1.2S. vasd contalnrnent High~S Red hdon Monitor

Pnmary Contahuncnt InlcprltyC4nnol bs nntav»d

> 20.000

Ik4 None

(

~ '

5

2.0 REACTOR PRESSURE VESSEL

~ ~

2.1.1

Parameter Opec,

2,2.1 None

Parameter

Visual observatcn ol a sustarncduncontrctted water level decrease

5 irl the Reactor Retuctrng Crnntyvnth ad dracaated tuel assemblesremamvtg covered by water.

2.1.2 2.2.2

I 23 Reactive Pressure Vessel Watslevel hoss!han of ertual lo low lowlow level I

A full reactor scree has beenInitiated with failure ofautonatlc trips to bring thereactor sub-critical

2.'13a 2.23

133 Reactor Pressure vessel dater 5-1st he43 level belou thc Top of Active

Fuel (TAF)

2.13 b

133. Reactor Pressure Vessel Water4,5 level canrct be determined.

A Iud reactor scram has beenin'tiated wdh Iadure ol Ixxhautomate and manual trips Io

1.2 bring the reactor subcfcical

AND

Strppress Ion Pool Teeperature > I ION

2.1.44

133. Reactor Pressure Vessel Vstcr43 level belier the Top of Active

Foci iTAF) for greater than15 ~ inutes

2.1.4 bh23. Reactor Pressure Vessel Water

4 level cannot be dctermrnedAND

Reactor Pressure Vessel RoodvtgPressure cannot be met vnthm 15tMlutes.

5-141 dr.for > 15

2 42 rrnrtabove Supp

Chamberpressure

1.2

2.2A

A iud reactor SCram has beeninnrated wim ladure ol bothautonatc and manual trurs lobnng the reactor sut>cntcal

AND

Suppression Pool Temperatute > 2OIVF

h,

3.0 PRIMARY CONTAINMENT

311

1.2. Drywea Pressure indrcanon > I.n pep3

AND

hcncanon d a leak geo

12,3

urudennrnd Or Preaaure BOundary o IP.leakage

> leonrn

or

Containnent IIydroyenCancan'teat'Ion

> sx

identified testate > SS Spo

3.1 2 322 None S.S.S None

1.2. Drywea pressure Inckcaoon > l,0pno3

AND

Indrsannn d n leak irnO

313 S.S.S None

Itaaid. unendowed decrease inDrywea pressure iooowinp insiaipressure rne aaoue 30 puy

123 ANO

Indcabon d a leak gsocnnrnerd

3.1 a 32,4 None S.S.S None

Oryuetl Pressure or Suppressloe12. ckanber erasers exceed > sspwo3 Contalnnest Nax lans latsrnal

predure

I~ [

4.0 SECONDARY CONTAINMENT Are 0De:o~ ice.rayo e us

Opec.Mode

4.1.1 None

ModeOper.

4.2.1

Oesccipdon ParameterTable 4.1.1

Secondary Containment Maximum Safe Values(Temperature)

SC AREA DESCRIPTIONMAXSAFE

AREA TEMP('F)

1.2. Vakd Main Steam Une High3 Radiation Trip

7 TimesFurr liower

Sacxttround

749 It el RWCU Penetrathn RoomRWCU Pump Room

212212

719 lt el Main Steam Tunnel

683 tt el RHR Pipe RoutingHPCI Pipe RoutingRCIC Pipe Routing

240240240

4.1.2 Hone 4.22645 lt el HPCI Eqeprnent Area

645 lt el RCIC Equipment Area

Valid MSIVClosure Signal orIndication

1,2,3 AND

Main Steam Une Row indication

4.13 4.23 A and (8 or C)

1.2,3

. An Unisolable Primary SystemLeak is drscharging into SecondaryContainment

AND

Secondary ContatneentTeepera'ture exceeds itaxttxssSafe Tesperature Llult Inany area of Table 4.l.l

4.1.4

An Unisotable Primary SystemLeak is discharging into SecondaryContainment

ANDSecondary Cental neentTecrperature exceeds HaxlsxssSafe Tecperature Lttrtt Inany area of Table 4.l.l

ANDAny parameter in Table 4.1.2.Indications ol Fuel CladdegDegcadatiOn, haS been exCeeded.

A. Main Steam Une Breakdischarging outside Primary andSecondary Containment

1.2. AND3 ILTurbine Building Main Steam

Tunnel Temperature aboveisolation Setpoinl

ORG Turbee Burtdeg SPING HighAlarm

4.2.4 A and (8 or C) and DA. Main Steam Une Break

dischargeg outside Primary andSecondary Containmenl

ANDB Turbine Building Main Steam

12, Tunnel Temperature aboveIsolation Setpoint

ORG Turbine Buitdeg SPINO High Alarm

ANDD. Any parameter in Table 4.1.2.

indications ol Fuel CladdegOegradaten, has been exceeded

Table 4.1,2

Indicallons ol Fuel Cladding Degrsdatlon

PARAMETERDESCRIPTION

Reactor Coolant ActivityDeterminedby Sample

) 300 uCilgrn[uCI/cc oc uCirmtlpose Equivalent

ladino.131

Valid Containment High Radiation Monltoc

Reactor Pressure Vessel Water Level M -161 inches

'lccrrcmsrr~ temprvaare exceeds 22$ ~F ~ 2$ cslv onecrign rackvmust be ercrred manrrsty ro rhe rixscsaon co orrsei ~ crownscare enor,

5.0 RADIOACTIVITYRELEASES ~o

DnCrphon Pvseerv Opec Dosclcx»n Tablt 5 ~ .I

I.I.I.~

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S.l.l.sTio curtcvttvo Il ~Ittt~

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CR

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10'et/el~

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Itch Sttc).I).2.1

ALL

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5.2 2

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Abnorea Rcljklf ties so Ior Qfsawlhsn 60 meuff4 witch exceeds 2hlvlos Technical Speed»a»nx

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Tech SpeclffI I

51112

Tt»ISprr.illI I~ns

21112

54.1

I/ntrpected lnpffe raha»nIovtffexceed IOO le»l Lsw

ALL Range Area f4ha»n VonnorSapout and levels 4I~ vtnl»d bysrxvoy

> 100 uroshuh les~olpcwShee e

Pscl Ouupport41s, their

5420ALL Any votd peaks Quslnl lath Rsntt

Arts Relate Voter

54.2nVdd uCS nldna eu»un SSsll n)cux

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Stlha Ne44'l'54,

424

Tlgh N 124

)cthe St10

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IIA

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Ruhr)

219 h ll 5'.

Id) hcl NA

SIS N CI

24I'5

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5051

455)5555NA

HKI2Oupete AresRCC Igrttsta ArtsRHft Peep Rcoe 0RHR Pussp Role AfS Step lkoe

IOIIOI

)CO IOI

104

tgv

IOI

Rrstr 001 N Star14404 0 I Nu evsvRtea ffr »I svav

Secondvy Containment Vaxhuum Sale Vaffes(fffd4tlon)

~ Dose caku41»ns art unsvaffbltsnd Iwoconsocuhn lO meuffav«age SPPIG vent meteorrtsdeQC kvs¹ao:

ALLNobffGss AeffutAae

ffdet.t)1 Asffsse Aao

Tccsl SseAOIO4so

> 2.2 4 ffvucvtne

> Sl 4 NIICCnse

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VWe Body 4

OR

CMd Thyrord

4 O.IRfer

rr ties ofnltsrt> O.SR

/srmrsat tl

rel ~ tse

L2.

AnUn¹dabff Pnrn«y SvsremLeak» etch«png ICCO Second«2

AND

Secosdsry Costal sheet Rsdlatloslevels txcted Iffxlem SafeRtdlatlos lani ls asy sna ofTable 5.4.)

Ttbff 542

Indlcstkms ol Fuel Cladding Tffgr«fadon

PARAVE'TEADESCRIPTION

Reactor coo4nl Acthdly Da«nwlodby SaeCff

> 000 OCugm[OCvrtc or OCurnflDose Equrvaonl

lodne 131

Vakl Contalnmenl Nigh R«$ 4tlon Vonhor > 50A/tu'1

~

Dose calcv4hons aro vesvacsbff«Id Iwo onstcvow 10 meuffsvtfsge SPING volr mentorrtsdegs ecscsca

ALLNcbffGss Rtffsse Aae

kvsho.t)1 AtffuaAsff

> 22 4 tsv4Cvee

> $2 4 1st4C»ne

5.2 4

Adust or fekecffd dose a orbryced Ihs EPB laN

ALLWhoffBody

OR

Chfd Thynxdt

r ltftr

hrrsllu ttre 1 tnt4 SRftr

ewsttu 41rc 1 eat

5.44An I/n»offbffPnn»ry Sysffeteak» ksch«geng kso Secondary

AND

Stcovtsry Coatatr»est Rsdlstloslevels excetd alxlmsa Safe244istloa Itvel la aay area afTable $ .4.)

AND

Anypa«net«m Table 5 42.hxscabcv» d Fud CffddegDOQI«fa»n, MS bOth trcttdtd.

Reactor Pressure Vessel V/ster Level 101 IOChss

4/Vvnry~ rwtuwmn e»rtev 2)SIP ~ 2$ rtrvcortehe rrrxvevr tu send OI«story 4 er edcstcn rt urrtr ~ Orwescsu tnw

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4 Ctultt44 Dost fotlvslttt sl tlt tlyrsl4 frts rseltl44lse.

6.0 LOSS OF POWER 40~wrsrv

0,1.1

Loss ol Power lcom StarsupALL Tcndooa»cs '104iid 20 lo ocher

una

> lsccalutes

Unptanned S»s d veal DC Power

4 atdctated by ndercdt404 On Odd 1 co~and Dgts 12$VDC cnam

lsc > ts

dntrecvdon panel bc»ehTatee 62

125V DC MAINOISTIUBUTION PANEL BUSES

012bLoss d Power teem StacsupTrsaacra»rs 10 and 20 al eel»r une

ANO04.

As 4 16Kv Ess Buses on othet une4» enegclied

> 1$cnvvncs

01.24toss el Power lrom stwsupbanslormers 10 and 20 lo saner une

Ig,3 AND

asnutss> 1$

Dnsee AC Power Is reduced as asingle 416Kv Ess Bus on saner une

0,2.2

Loss d veal DC power bvos aced byIg, cmdenoa404 ori D6N arid D0243 12$v oc cnaav chstnbcecon pand

buses

UNIT IID61 ~ID615ID624ID625ID634I0635ID6llID6l5

UNIT 22D61 ~

20615206242D62$

2063$2064 l2D645

0.'13 023

Loss d powec Ircm StarsupTr nsloa»rs 10 and 20 lo ecd»rucvt

3 AND

AU.4.1d KVESS Buses on sanerucvt sre c»c0gtsd

> IScnlnvtc\

Loss d ved DC Power intscated byIg. ~ On ALL0» t2$V OC 4

3 cnain ctrevaem pand busestSee T44te 02t

0,1» 02A None

LOSS Ol Puwer anm StarHPbanskca»cs 10 and 20 lo ochertv»

ANO

ALL4 16 KVESS Buses on eeheclaic ace eee»cYQcted

> 4 hours

TAl INTERNALEVENTS

I

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ccI»V

aw»w »ace»aw»cwNIt cww»wawaacwc»1 Isaac»wc a»»twsw

~e c»aawsCOlwwar»a

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CSc Loaesacs feel Ia»ftt~tcsca»cw ass»a Cosa»

a w»gva

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IIICit»v NNwcwwNVW~sstoca sso lass tet

wtSWNVCS

0 a»caaaIMIVN V~cl»l sr»w aossaewNwca»ta ~ sewcw»av sttssww csocsW SNtwtIII w »wow

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Tae» 1 I1

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I Ias es aw «oe sowr awe» wwaanvc»w a at aaa l»WINtINN»N

lotlo 1.5.1

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Col ~ CIISNIIISCS . Sett Sto TITNI

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fto Iaw

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8 0 EXTERNAL EVENTS

IRMI)ISI

Cf)

ALL

8 I I

Parafnefw

CNIirwl«corks lefral lifKIMlewrl14 Ksl4a es swicelel efi~ f leftal lakfoal4s, 1stctwferwas tl loire Me» Io ~~rNA» owrfie IMietwletlewaeiitl M Ma-ekal Kroi tl tee tlNI,Nke li receiwl rna ~ ttll4ltsorct lt t. ltfenl. Slalt, M IKalI~ N«fcwet wwlts MCcwfralt 5tcwkll

kl

C LI a

lire ~kelt ~ sloe silk SinlCINCMica Iss ~ffKlel~INI IfntMlnelnl 4 ~KIII IN M ellol~ Ie«» NKKM,

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&&IcHcpcrt bf Coal. Cfctcyor &fatcfsceh for

ALL pcccrad fvsfcccc¹ 01 scc pclscncclblscdcn cgscc leans

OpecMode

8314

Ownage lo soccer pfwa areaALL resuhsng From a

vohrle assn GR Ecpa¹son

nncf 4 hlk C Iar nc an ol luk cress

831b

Twbr»»euro losuaang ul cakAQ13 pw»uaffon iu dki»QO IO TurtwN

Genera»r Sea»

832

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Erthfbref» de»aed by swsmso4stlurnonfsf NIV»hay pwsonnef

8.~ 141

Roport ol a l«r»do 5lflung ikdwlIINPROtECTEO AREA

&4 24Esrthouaae gresfw 8»n Operasng &OSQSac¹ Ertrthfaraae (OBE)

Tsb» 82

Plant Areas

conffa sffuaweONSOI Genwaa«&uecangc

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'esaccBuddngMOIATlenslofaNf Area

S»rlup TfwickriiielAreas

Cg

111

IMIL

IUICO

Cwii«el swell ~ 4lfwi« tr Kl 4 IMsisal tkiflClls¹fc 4 lelkal« ori~ Oee of oorr ofe« sensa Iai

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~ 5erreHkkes allerlsl ot sisalMeieaMI M tteltal efceaerisi~Nfallel lt tie Ilia owfaIIM~f IMNMtl 14 nKIeri,

813

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Report of wiye»af» Sflua«aALL damage on any 01 SN p»nc VITAL

STRUCFURES'esuhng IroncVefiae crash ORM¹tde unpsa OR

Esp»sion

'nelor» yac» as Iv»c 01 pine vc¹Smrsues

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)717) 4578377

&43 None

Tab» 83

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Everss are in pmcess a tmveoccurred wtxch bxac¹t 4 pomraNIdepradabcn Ol ON level Ol Sat¹y dON Planl. NO r¹eases dmdcactve mateo¹ leehxnnQcdsrta mspons4 cfnlcncollnQ areexpected unless lunherdeQr4d4bon d s¹ety sysmnis

Evems slain process or haveocclÃled whictlbwotw 4ilachml orpctenb¹ suboandat SQad4bon ol

ALL ON leo¹ ol salety ol oN ctanL Anyr¹eases 4lt expected lo ba ante«Ilo slnaa Ilactnns ol the EpAProtecbw Acbon Grnd¹rNexposure levA.

~ EW¹S are in paean a haveoccurred which Involve altu4I crlaeiy mayar taaurea ot pierstunaens needed Ice pret«bond

AtL O» putrbC. Any releases are ndaxp«ted lo rasun in exposureIev¹s wtiioh exCeed EPAPmtectne Acbon G¹d¹ir»exposure Iev¹s except near ONsrtt boundary.

Events art in process or haveoccurred wtuch invctva «ruat orrwnANilsubstaro4I cof4

deQladabon of mebxvQ wiolALL pxenbal lor tes d corasmmera

bttQnty. tie!eases can tNreasonably expected lo exceedEPA Pret«twe Arson Guideaneexposure lewiS Otisae lor moreOlail IIN VlitlbattSIte 4ft4.

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ATTACHMENTTO PLA%032

General No Initiating Conditions (/Cs)

The SSES EALs are not grouped under initiating conditions (ICs) as is specified inNESP-007. Per NESP-007, ICs are "one of a predetermined subset of nuclear powerplant conditions where either the potential exists for a radiological emergency, or suchan emergency has occurred." EALs are "a predetermined, site-specific, observablethreshold for a plant initiating condition that places the plant in a given emergencyclass." Although EALs are not required to be grouped under ICs by the regulations, theuse of ICs is advantageous from a human factors perspective. Grouping EALs underICs willindicate to those who must use the EALs how an EAL (or several diverse EALs). ~

is related to the plant condition which is of concern. This will assist the emergencydirector in the use of judgement in making the correct event classification. The lackof ICs for loss of fission product barriers is of particular concern to the staff. SSESshould include ICs with their EALs where appropriate.

R~es ense

";, The philosophy, for handling emergencies at PPSL is symptom based not eventbased. This philosophy is reflected in our Emergency Operating Procedures(EOP) and therefore the EALs would naturally follow the same philosophy. Byassessing the symptoms, the causes for the symptoms are identified and thebig picture is obtained. Having ICs would slow the process established forclassification. The Emergency Directors are senior plant staff members whicheither have SRO License or are SRO certified. The EDs have many years ofexperience in nuclear plant operation. They are trained to assess the wholesituation and not dwell on one particular symptom but by treating thesymptoms the correct classification of the event is obtained.

The loss of barrier matrix as described in the NUMARC methodology has beenincorporated into the proposed EALs. The following additional guidance hasbeen added to Section 3.1 of the Guidelines for Using Emergency ClassificationProcedure:

To the extent possible this procedure uses a symptomatic approach toidentifying emergency conditions. This approach complements the EmergencyOperating Procedures (EOPs) and is intended to be used in concert with theEOPs when responding to an emergency. These EALs are based on theguidance provided in "NUMARC NESP 007 Methodology for Development ofEmergency Actions". One key section of this document addresses FissionProduct Barrier Degradation. The criteria provided follows:

~ Unusual Event - Primary Containment fission product barrier is lostor potentially lost.

~ Alert - Either Fuel Cladding or Reactor Coolant Systemfission product barrier is lost orpotentially lost.

~ Site Area Emergency - Any two fission product barriers are lost ofpotentially lost.

~ General Emergency - Any two fission product barriers are lost and thethird is lost or potentially lost.

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ATTACHMENTTO PLA-4032

NESP 007 defines what is meant by "loss" and "potential loss" of each fissionproduct barrier. The attached EALs implement this approach but the symptombased format used may mask this relationship. Therefore when assessingemergency conditions the status of the fission product barriers should beconsidered extremely important. Ifone barrier is breached an emergency shouldbe declared at either the Unusual Event or Alert Level. If two barriers arebreached a Site Area Emergency should be declared. If two barriers arebreached and the third is breached or seriously threatened a General Emergencyshould be declared.

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ATTACHMENTTO PLA&032

EAL 1. 1. 1.a Chemistry determines that reactor coolant activity exceeds technicalspecification limits (Tech Spec 3.4s5 4 pCi/gm dose equivalent /-131

This EALdoes not include the technical specification coolant activity limitof 100/E-Baras a condition and therefore deviates from NESP-007, IC SU4, EAL 2. This EAL shouldbe revised to include the IOO/E-BAR coolant activity limit or additional informationshould be provided to justify this deviation.

This EAL also deviates from NESP-007, IC SU4 in that Mode 5 is not included as anapplicable mode. This EAL should be revised to include Mode 5 as an applicable mode. ~

or additional information should be provided to justify this deviation.

~Res ense

This EAL has been revised to include the 100/E-BAR coolant activity limit.

With regard to not including Mode 5 in this EAL, Technical Specification 3.4.5's not applicable in Mode 5 and therefore, there are no Technical Specificationlimits in Mode 5. In order to avoid confusion for the operators, this EAL hasbeen written to be consistent with the Technical Specifications.

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ATTACHMENTTO PLAR032

EAL 1.1.3 Reactor coolant activity determined by sample () 2000 pCilgm doseequivalent 1-130

This EAI is classified as a Site Area Emergency. This EAL deviates from NESP-007,IC FG1 "loss of any two barriers and potential loss of third barrier" which is classifiedas a General Emergency. According to the basis for the SSES EAL, coolant activity atthis level is indicative of 20% clad failure and is also indicative of a loss of the RCSbarrier.

In the basis for the NESP-007 guidance for this EAL, it is stated that "regardless of..whether containment is challenged, this amount of activity in containment, ifreleased,could have such severe consequences that it is prudent to treat this as a potential lossof containment such that a General Emergency declaration is warranted. "

This EALshould be revised to correspond to the General Emergency classification levelor additional information should be provided to justify this deviation.

~Res ense

This EAL is strictly for loss of the fuel barrier in the order of 20% clad failureand does not imply that the RCS barrier has been lost or that containmentintegrity is being challenged. If fuel damage was this severe, a site areaemergency is warranted. This EAL is being used as a progression for coolantactivity and does not meet the criteria of NESP-007 FS; therefore the referencein the Technical Bases to NESP-007 FS has been deleted.

This EAL could be deleted since NESP-007 FS is met by the EALs onContainment Radiation monitors and reactor water level. We have elected toretain this EAL since it is a logical progression from an alert condition forreactor coolant activity.

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EAL 2.1.1 Visual observation of a sustained, uncontrolled water level decrease inthe reactor refueling cavity with allirradiated fuel assemblies remainingcovered by water

This EAL deviates from NESP-007, IC AA2, EAL 3 in that "defueled" is not includedas an applicable mode. This EAL should be revised to include the defueled mode asan applicable mode or additional information should be provided to justify thisdeviation.

R~es ense

In the defueled mode as defined and controlled at Susquehanna SES, all fuelassemblies have been removed from the reactor cavity and placed into thespent fuel pool. Therefore, the defueled mode is not included in this EAL. Theloss of inventory from the spent fuel pool in all modes is covered by EAL 1.3.1.

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EAL 2.1.4.a Reactor pressure vessel water level cannot be maintained above the topof active fuel for ) 20 minutes

This EAt: appears to deviate from NESP-007, IC, FG1, EAL "Reactor vessel water levelless than (site specific) value and the maximum core uncovery time limit is in theUNSAFE region." From the supporting information provided with the SSES EAL it couldnot be determined whether the 20 minute time limitspecified in the EAL correspondedto the NESP-007 EAL condition of the maximum core uncovery time limit being in theUNSAFE region.

Additional information should be provided regarding the relationship of the 20 minutesspecified in the SSES EAL to the maximum core uncovery time limit being in theUNSAFE region specified in the NESP-007 EAL. If the SSES EAL does not correspondto the NESP-007 EAL condition of the maximum core uncovery time limit being in theUNSAFE region, then the SSES EAL should be revised to be consistent with theNESP-007 EAL or additional information should be provided to justify the deviation.

~Res ense

This EAL has not been equated to the maximum core uncovery time limit'sUNSAFE region in the EOPs. As part of the IPE analysis, a computer analysisof core damage resulting from design basis events was performed. As part ofthis analysis one of the specific scenarios was for the worst case condition ofhaving no water injected into the vessel during a break in the recirculation line.This analysis showed that the first release of fission gas was at approximately12 minutes with the first evidence of core melt at approximately 24 minutes.The analysis also showed that the time to first release of fission gas increaseswith an increase in water being injected into the vessel. There is virtually nofission gas release if the minimum number of ECCS pumps are injecting to thevessel,

Based on this computer analysis a revised time frame of 15 minutes has beenselected as a reasonable time based upon all the analysis performed to allow forrecovery of water level prior to significant fission gas release. If the operatorhad no method of injecting into the vessel, a General Emergency would bedeclared prior to 15 minutes based on EAL 9.1.4 and training. Also if the fuelwere to be extensively damaged prior to 15 minutes, the containment radiationmonitor would see an increase and detect this damage.

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EAL 2.1.4.b Reactor pressure vessel water level cannot be determined and reactorpressure vessel flooding pressure cannot be met within 20 minutes

This EAZ appears to deviate from NESP-007, IC FG1, EAL "Reactor vessel water levelless than (site specific) value and the maximum core uncovery time limit is in theUNSAFE region."

From the supporting information provided with the SSES EAL, it could not bedetermined whether the 20-minute time limitspecified in the EAL corresponded to theNESP-007 EAL condition of the maximum core uncovery time limit being in the.-UNSAFE region.

Additional information should be provided regarding the relationship of the 20 minutesspecified in the SSES EAL to the maximum core uncovery time limit being in theUNSAFE region specified in the NESP-007 EAL. If the SSES EAL does not correspondto the maximum core uncovery time limit being in the UNSAFE region then the SSESEAL should be revised to be consistent with the NESP-007 EAL or additional

"information should be provided to justify the deviation.

~Res ense

A revised 15 minute time frame has been chosen based upon our IPE computeranalysis and in order to allow operators to perform the actions required toinitiate core flooding. As long as there is a water source the core damagewould not be expected to occur within the 15 minute time frame. See theresponse to Item 5 above.

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7. EAL 2.2.1 A fullreactor scram has beeninitiated with failure ofautomatic trips tobring the reactor subcritical

This EAL deviates from NESP-007, IC SA2, "Failure of Reactor Protection SystemInstrumentation to Complete or Initiate an Automatic Reactor Scram Once a ReactorProtection System Setpoint Has Been Exceeded and Manual Scram was Successful,"in that SSES classifies this event as an Unusual Event while the NESP-007 guidanceclassifies this as an Alert.

As stated in NESP-007, the failure of the RPS to automatically scram the reactor is.more than a potential degradation of a safety system in that a front line automaticprotection system did not function in response to a plant transient. Thus plant safetyhas been compromised and design limits of the fuel may have been exceeded. AnAlert is appropriate because conditions exist that lead to a potential loss of the fuelclad or reactor coolant system.

This EAL should be revised so that it is classified as an Alert or additional informationshould be provided to justify this deviation.

~Res ense

Based upon the agreements from the meeting on April 7, 1993 betweenNUMARC and NRC, this EAL has been deleted.

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8. EAL 2.2.2 A fullreactor scram has beeninitiated with failure ofboth automatic andmanual trips to bring the reactor subcritical

SSES classifies the failure of the automatic and manual reactor protection system asan Alert if reactor power is less than 5 percent following initiation of the automatic andmanual scram and as a Site Area Emergency if power is greater than 5 percentfollowing the scram.

In NESP-007 IC SA2 and SS2, the failure of automatic RPS with a successful manual.scram is classified as an Alert and the failure of both the automatic and manual RPSscram is classified as an Site Area Emergency. A successful manual scram is any setof actions by the operator(s) at the reactor control console which causes control rodsto be rapidly inserted into the core and brings the reactor subcritical. If the reactor isnot subcritical following the manual scram, the event should be classified as a SiteArea Emergency.

This.EAL.should be modified to include the condition that the manual scram wassuccessful in bringing the reactor subcritical or additional information should beprovided which justifies including a reactor power level after the scram instead of a

successful manual scram. This justification should provide the technical basis for themaximum power level specified in the SSES EAL as the delineation between the Alertand Site Area Emergency, including (1) the capability of safety systems to remove theheat produced from the specified reactor power level plus 100 percent decay heat(with containment isolation), (2) the expected operator actions for this type of an eventas guided by the Emergency Operating Procedures, and (3) the plant condition andcorresponding EAL which will result in the escalation to the Site Area Emergencyclassification level.

~Res ense

Based upon the agreements reached during the April7, 1993 meeting betweenNUMARC and NRC, this EAL is revised to read as follows:

A full reactor scram has been initiated with failure of automatictrips to bring the reactor subcritical.

With this revision, this EALwillbe consistent with NESP-007, IC SA2 howeverwe have elected not to include the successful manual scram. It is our opinionthat including the successful manual scram adds confusion and is inconsistentwith our convention of not stating positive conditions. The intent of this EALis to declare an emergency on the loss of an automatic scram even if themanual scram is successful. When the successful manual scram is added, itdoes nothing to enhance the EAL and adds amount of uncertainty on why anAlert is being declared since the reactor is in a safe condition (subcritical). We,therefore, have not included the successful manual scram in this EAL.

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ATTACHMENTTO PLA<032

9. EAL 2.1.3.b RPV water level cannot be determined

Mode 5 is not included as an applicable mode for this EAL. Although the NESP-007does not specify applicable modes for this EAL, additional information is needed todetermine whether or not mode 5 should be included as an applicable mode for theSSES EAL.

R~es ense

This EAL has been revised to include Mode 5.

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10. EAL 2.2.3 A fullreactor scram has beeninitiated with failure ofboth automatic andmanual trips to bring the reactor subcritical, and reactor Power ) 5percent

This EALdeviates from NESP-007 IC SS2 in that the SSES EAL specifies that "ReactorPower is ) 5 Percent" while the NESP-007 EAL specifies that "the manual scram wasnot successful." For the reasons discussed under EAL 2.2.2 above, the condition of"Reactor Power ) 5 Percent," should to be changed to "manual scram was notsuccessful" for EAL 2.2.3 or additional technical justification for including the powerlevel in the EAL should be provided.

~Res ense

This EAL has been revised as follows:

"2.2.3 A

A full reactor scram has been initiated with failure of both automatic andmanual trips to bring the reactor sub-critical

'ND

Suppression Pool Temperature ) 110'F

With this revision, this EAL will be consistent with NESP-007, IC SS2. Theparameter of suppression pool temperature is used as confirmation that thedecay heat removal capability is being challenged as stated in the basis forNESP-007 IC SS2. The suppression pool temperature parameter would indicatethat the main steamline isolation valves are closed and the reactor is producingmore heat than the decay heat removal systems can handle.

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11. EAL 2.2.4 A fullreactorscram has beeninitiated with failure ofboth automatic andmanual trips to bring the reactor subcritical, and reactor power > 5percent, and suppression pool temperature > 200'i=

This EAL deviates from NESP-007 IC SG2 in that the SSES EALspecifies that "ReactorPower is > 5 Percent," while the NESP-007 EAL specifies that "the manual scramwas not successful," For the reasons discussed under EAL 2.2.2 above, the conditionof "Reactor Power > 5 Percent," should be changed to "manual scram was notsuccessful" for EAL 2.2.4 or additional technical justification for including the powerlevel in the EAL should be provided.

In addition, this EAL specified "Suppression Pool Temperature > 200'F" as the soleindication of the corresponding NESP-007 EAL conditions that "core cooling isextremely challenged" or "indication exists that heat removal is extremely challenged."Additional information should be provided which justifies not including other sitespecific indications for the corresponding NESP-007 EAL conditions.

R~es ense

In this EAL, we state that the reactor is not brought subcritical by both theautomatic and manual trips. This is our way of saying that both the automaticand manual trips were unsuccessful. The power level has been deleted. Thesuppression pool temperature is used to show that core cooling capability isbeing extremely challenged. Other indications of core cooling and heat removalbeing challenged would be water level indications and loss of systems toprovide water to the core or remove heat from the suppression pool. The waterlevel indications are covered in EALs 2.1.1 thru 2.1.4. The loss of systems iscovered in EAL 7.3.3.

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12. EAL 3.1.1 Drywellpressure indication ) 1. 72 psig and indication of a leak intocontainment

This SSES EAL is classified as an Unusual Event. This EAL deviates from NESP-007IC FA1, which is classified as an Alert.

In the basis for the NESP-007 EAL it is stated that "the drywell pressure setpointshould be based on the drywell high pressure alarm set point and indicates a LOCA,A higher value may be used if supporting documentation is provided which indicatesthe chosen value is less than the pressure which would be reached for a 50 gpm RCS, ~

leak." The documentation provided by SSES supporting this EAL indicates that 1.72psig will be reached if there is a 50 gpm leak. It is appropriate that an Alert bedeclared for a leak of this magnitude.

This EAL should be modified to correspond to the Alert level classification or additionalinformation should be provided to justify this deviation.

R~es ense

This EAL does not deviate from NESP-007 IC FA1. As stated in the bases forNESP-007 IC FA1, the pressure to be used is a pressure which is less than thepressure which would be reached for a 50 gpm Reactor Coolant System leak,Therefore, the bases for a declaration of an alert is a reactor coolant systemleak. NESP-007 IC FA1 explicitly allows the use of a pressure of greater thanthe drywall high pressure alarm setpoint if supporting documentation isprovided which indicates the chosen value is less than the pressure whichwould be reached for a 50 gpm reactor coolant system leak. As shown in theanalysis which was provided, the maximum drywell pressure for a 50 gpm leakis greater than 4.0 psig. Therefore, the selection of 3.0 psig for theclassification of an alert as stated in our EAL 3.1.2 is in accordance with thebasis for NESP-007 IC FA1.

NESP-007 does not have an EAL for the classification of an Unusual Event ondrywell pressure. We have chosen to include this EAL based on the drywellpressure of 1.72 psig resulting from a reactor coolant system leak as anUnusual Event since it is a precursor to the loss of reactor coolant system.

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13. EAL 3.1.2 Drywellpressure indication ) 3. 0 psig and indication of a leak intocontainment

As disc6ssed for EAL 3.1.1, NESP-007 states that "the drywell pressure setpoint forthis EAL should be based on the drywell high pressure alarm set point and indicatesa LOCA. A higher value may be used if supporting documentation is provided whichindicates the chosen value is less than the pressure which would be reached for a 50gpm RCS leak." The documentation provided by SSES supporting this EAL indicatesthat 1.72 psig willbe reached ifthere is a 50 gpm leak. Therefore the setpoint for thisEAL should be changed to 1.72 psig.

This EAL should be modified so that the "Drywell Pressure Indication" limit is changedto 1.72 psig or additional information should be provided to justify this deviation.

~Res ense

This EAL does not deviate from NESP-007 IC FA1 since the use of a pressurevalue-greater than the isolation setpoint is allowed by the basis of NESP-007IC FA1 as long as the pressure value selected is less than the pressureassociated with a 50 gpm reactor coolant system leak. We have chosen toselect this option.

See our response to Comment ¹12.

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ATTACHMENTTO PLAX032

14. EAL 3.1.3 Rapid, unexplained decreasein dry wellpressure followinginitialpressurerise above 3. 0 psig andindication of a leakinto containment

As discussed for EAL 3.1.2, this EAL should be modified so that the "Drywell Pressureindication" limit is changed to 1.72 psig or additional information should be providedto justify this deviation.

R~es ense

Since 3 psig has been chosen to be the pressure value for the classification ofan alert, this EAL is consistent with that pressure. See the response toComments ¹12 and ¹13.

Page 15


15. EAL 3.1 4 Drywellpressureandsuppressionchamberpressureexceedcontainmentmaximum internal pressure

Events which result in Drywell or suppression chamber pressures of this magnitude areindicative of a loss of RCS, potential loss of fuel clad and potential loss ofcontainment. NESP-007 does not contain an EAL for high drywell pressure which isindicative of a condition of potential loss of fuel clad. This deviation is acceptablesince the high drywell pressure is indicative of a major loss of RCS which couldpotentially damage the fuel clad. Consistent with this conclusion, either high drywellpressure or high suppression chamber pressure should result in the same classification..

Therefore, this EAL should be modified to indicate that either drywell pressure orsuppression chamber pressure exceeding containment maximum internal pressure willresult in an Alert classification.

R~es ense

„This'AL .has been revised to indicate that either drywell pressure orsuppression chamber pressure exceeding containment maximum internalpressure will result in a General Emergency classification.

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16. EAL 3.2.1 Unidentified or pressure boundary leakage ) 10 gpm

This EAL deviates from NESP-007, IC SU5, EAL 2, "Identified Leakage Greater than25 gpn1,"'n that identified leakage is not included in this EAL.

The licensee's basis for not having an EAL for identified leakage is that the leakagecomes from known, monitored locations such as valve packing and pump seals whichdo not represent a "leak before break" concern. The NESP-007 IC for RCS leakagespecifically included an EAL for identified leakage. In the basis for this IC it is statedthat the RCS leakage IC is included as an Unusual Event because it may be a precursor. ~

of more serious conditions and, as result, is considered to be a potential degradationof the level of the safety of the plant.

This EAL should be modified to include the condition of "identified leakage greater than25 gpm" or additional justification for not including identified leakage should beprowded.

R~es ense

This EAL has been revised to include an identified leakage of greater than 25gpm.

Page 17

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ATTACHMENTTO PLAQ032

17. EAL4.1s2 An unisolable primary system leak is discharging into secondarycontainmentandsecondary containment temperature exceedsmaximumsafe temperature limitinone area of 7able 4.1.1

This EAL deviates from one of the example EALs for NESP-007, IC FS1, i.e. "unisolableprimary system leakage outside drywell as indicated by area temp or area rad alarm."The NESP-007 EAL is indicative of a loss of the RCS and Primary Containment fissionproduct barriers and the condition is classified as a Site Area Emergency. In contrastto this, EAL 4.1.2. is classified as an Alert.

In the basis for this EAL the licensee states that this EAL represents a challenge toboth the RCS and containment fission product barriers. EAL 4.1.3, "An UnisolablePrimary System Leak is discharging into Secondary Containment and SecondaryContainment Temperature exceeds Maximum Safe Temperature Limit in MORE thanone area of Table 4. 1. 1, " is classified as a Site Area Emergency and in the basis itis stated that this EAL is indicative of a potential loss of RCS and loss of containmentfission product barriers.

The licensee uses the indication of more than one area exceeding the safe temperaturelimitversus one area exceeding the safe temperature limitas a discriminator on the sizeof the leak,

This EAL should be revised to correspond to the Site Area Emergency classificationlevel or information should be provided as to the size of the leak which would resultin one area temperature exceeding the maximum safe area temperature and the sizeof the leak which would result in more than one area and to further justify thisdeviation from the NESP-007 guidance.

~Res esse

This EAL has been deleted.

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18. EAL 4.1.3 An unisolable primary system, leak is discharging into secondarycontainmentandsecondarycontainment temperature exceeds maximumsafe temperature limitinMORE than one area of Table 4. 1. 1

For the reasons discussed under EAL 4.1.2, this EAL should be revised to change thecondition that "MORE than one area... " to "any area" or information should beprovided as to the size of the leak which would result in one area temperatureexceeding the maximum safe area temperature and the size of the leak which wouldresult in more than one area exceeding the safe temperature limit and to further justifythis deviation from the NESP-007 guidance.

~Res ense

This EAL has been revised to change the condition that "MORE than onearea... " to "any area...".

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19. EAL 4.1.4 An unisolable primary system leak is discharging into secondarycontainmentandsecondary containment temperature exceedsmaximumsafe temperature limitin MORE than one area of Table 4. 1. 1 and anyparameterin Table 4.1.2, "Indication of Fuel Cladding Degradation, "

has been exceeded

The condition that the "Maximum Safe Temperature Limitis exceeded" is unnecessaryin this EAL, since there is sufficient indication that all three fission product barriershave been lost without including this condition.

The condition that the Maximum Safe Temperature Limitis exceeded should be deletedor additional information which justifies including this condition should be provided.

R~es ense

This EAL has been revised to change the condition that "MORE than onearea... " to "any area... ". The Maximum Safe Temperature Limit is usedas a confirmation that there is an unisolable leak. It is also used to maintainconsistency when escalating from a Site Area Emergency to a GeneralEmergency.

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20. EAL 5.1.3 Dose calculations are unavailable and tvvo consecutive 10 minuteaverage SPING vent monitor readingsindicate: Noble gas release rate )2. 7 x 10'Ci/min or iodine-131 release rate > 5.3 x 10'Cilmin

In EALs 5.1.1 and 5.1.2, the particulate release rate is included as an indication ofradioactive releases. Additional information should be provided which justifies notincluding the condition that, the Particulate Release rate is above a (site specific) setpoint, as part of this EAL.

~Res ense

Site Area Emergency and General Emergency Classifications do not useparticulate release rate as a classification trigger. Particulate effluent is notaddressed in Reg. Guide 1.3 or SRP 15.6.5. Reg. Guide 1.97 Table 1 specifiesmonitors which detect noble gas effluent across a wide range to detect actualbreaches, accomplish mitigation and verification. AtSusquehanna SES the postaccident monitoring instrumentation uses only noble gas monitors in accordancewith the guidance of Regulatory Guide 1.97 and NUREG-0578.

EPA and NUMARC recommend calculation of committed effective dose equivalent,which entails not only the external dose contribution (resulting from noble gasimmersion), but also weighted dose equivalent contributions to the total body frominhaled particulate and radioiodine. The EALs for Site Area Emergency and GeneralEmergency for noble gas have been calculated to include the weighted dosecontribution from inhaled particulate and radioiodine, accompanying external doseeffects of noble gases in the source term. The source term was developed based ondesign basis LOCA, which fission product inventory taken from NUREG-1228. Bothmonitored and leakage effluent pathways were considered in determining the offsitedose and the resulting EALs. The inclusion of radioiodine and particulate in the sourceterm of the calculation reduces the noble gas Site Area Emergency and GeneralEmergency Action Levels below what they would be ifthey were based only on effectsfrom immersion in noble gas. The EAL for particulate has not been separatelycalculated, first, because effects of particulate (and radioiodine) on the total effectivedose equivalent from a serious accidental release have already been included in theEAL for noble gas; second, the use of the noble gas EAL as the primary trigger for theSite Area Emergency and General Emergency Classifications is consistent withguidance in Reg.Guide 1.3, Reg. Guide 1.97, and with the SRP.

In addition EALs 5.1.1 and 5.1.2 have been revised to remove the particulate readings.

Page 21


21. EAL 6.2.3 Actual orprojected dose at orbeyond the EPB whole body ) 100 mremor child thyroid ) 500 mrem

The definition of the EPB (emergency planning boundary) was not specified in the EALbasis. Additional information should be provided concerning the definition of the EPB

to determine whether this EAL meets NESP-007 IC AS1.

~Res ense

The EPB is equivalent to the site boundary. This is consistent with theboundary as defined in the bases for NESP-007 IC AS1.

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22. EAL 5.4.2.c An unisolable primary system leak is discharging into secondarycontainment and secondary containment radiation levels exceedmaximum safe radiation level in one area of Table 5.4. 1

This EAL deviates from one of the example EALs for NESP-007, IC FS1, i.ess of"unisolable primary system leakage outside drywell as indicated by area temp or arearad alarm." The NESP-007 EAL is indicative of a loss of the RCS and PrimaryContainment fission product barriers and the condition is classified as a Site AreaEmergency. In contrast to this, EAL 5A.2.c is classified as an Alert.

In the basis for EAL 5A.2.c, it is stated that this EAL represents a challenge to boththe RCS and containment fission product barriers. EAL 5.4.3, "An Unisolable PrimarySystem Leak is discharging into Secondary Containment and Secondary ContainmentRadiation level exceed Maximum Safe Radiation level in MORE than one area of Table5A.1," is classified as a Site Area Emergency and in the basis it is stated that this EALis indicative of a potential loss of RCS and loss of containment fission product barriers.

The licensee uses the indication of more than one area exceeding the safe radiationlevel versus one area exceeding the safe radiation level as a discriminator on the sizeof the leak.

This EAL should be revised to change the condition that "MORE than one area..." to" any area..." or information should be provided as to the size of the leak which wouldresult in the one area radiation level exceeding the maximum safe radiation level andthe size of the leak which would result in more than one area exceeding the maximumsafe radiation level and to further justify this deviation from the NESP-007 guidance.

R~es ense

This EAL has been deleted.

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23. EAL 5A.3 An unisolable primary system leak is discharging into secondarycontainment and secondary containment radiation level exceedsmaximum safe radiation levelin MORE than one area of Table 5.4. 1

For the reasons discussed in EAL 5.4.2.c, this EAL should be revised to change thecondition that "MORE than one area... " to "any area" or information should beprovided as to the size of the leak which would result in the one area radiation levelexceeding the maximum safe area radiation level and the size of the leak which wouldresult in more than one area exceeding the maximum safe radiation level and to furtherjustify this deviation from the NESP-007 guidance.

R~es ense

This EAL has been revised to change the condition that "MORE than onearea ~ .. " to "any area . ~ . ".

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24. EAL 5.4.4 An unisolable primary system leak is discharging into secondarycontainment and secondary containment radiation level exceedsmaximum safe radiation levelin MORE than one area of Table 5.4. 1 andany parameterin Table 5.4.2, "Indication ofFuel Cladding Degradation, "

has been exceeded

The condition that the Maximum Safe Radiation level is exceeded is unnecessary in thisEAL, since there is sufficient indication that all three fission product barriers have beenlost without including this condition.

The condition that the Maximum Safe Radiation level is exceeded should be deletedor additional information which justifies including this condition should be provided.

R~es ense

This EAL has been revised to change the condition that "MORE than onearea ~ .. " to "any area ~ .. ". The radiation levels are used as a confirmation.that there is an.unisolable leak. It is also used to maintain consistency whenescalating from a Site Area Emergency to a General Emergency.

Page 26

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25. EAL 6.1.3 Loss ofpower from start-up transformers 10 and 20 to either unit andALL 4. 16 KVESS buses on either unit are de-energized > 15 minutes

In the basis for NESP-007 IC SS1, it is stated that time duration for the loss of powershould be selected to exclude transient or momentary power losses, but should notexceed 15 minutes. In the basis, for EAL 6.1.3 it is stated that 15 minutes has beenselected to exclude transient or momentary power losses. Further information shouldbe provided to justify why 15 minutes is appropriate for SSES.

R~es ense

The selection of 15 minutes allows the operator sufficient time to attempt torestore power to either the start-up transformers or the 4.16 Kv ESS busesprior to escalating the emergency classification. This time also allows forpower transients or momentary power losses. As part of the analysis forStation Blackout, the plant could cope with the total loss of AC power for 8hours with no adverse effects. Therefore, 15 minutes is not an untenablelength of time,to be without AC power.

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26. EAL 7.3.2.a Uncontrolled reactor coolant temperature increase (>200'F), andIsuppression pool temperature > 120'F or suppression pool level < 12ftJ

This EAL deviates from NESP-007, IC SA3, "inability to maintain plant in coldshutdown." One of the EALs for this NESP-007 IC is; "Loss of (site-specific TechnicalSpecification required function to maintain cold shutdown." SSES EAL 7.3.2.aspecifies the condition of high suppression pool temperature and low suppression poollevel as indications for the loss of Technical Specification required functions. Thesespecific conditions may not be the only indications or earliest site specific indications..of this condition.

This EAL should be modified to include additional indications, such as loss of RHR, asmay be appropriate for SSES.

~Res ense

The Technical Specifications for Susquehanna SES require one RHR pump anda flow path to be available for decay heat removal. The TechnicalSpecifications also allow alternative methods (unspecified in the TechnicalSpecifications) to be used in removing decay heat. These alternative methodsinclude venting water to the suppression pool through the SRVs and making upwith condensate or core spray. Also the Reactor Water Cleanup System canbe used to remove decay heat. The listing of loss of all these systems (andothers) seems to be unnecessary because it adds complexity to plantoperations. Also to declare an Alert on the loss of RHR alone would be overlyconservative and cause unnecessary public reaction and anxiety. PP&L hasseen this type of reaction when we declared an Alert for the loss of shutdowncooling. The key factors in the inability to maintain the plant in cold shutdownare whether or not there is core cooling and whether or not decay heat can beremoved. These factors are best indicated by the measurement of reactorcoolant temperature and the ability of the suppression pool to act as a heatsink.

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27. EAL 8.1.1 Confirmed security threat directed toward the station~ ~ ~

This EAL deviates from NESP-007, IC HU4, in that site specific safeguard events arenot delineated in the EAL, i.e. the SSES EAL does not indicate what constitutes a

"security threat." The basis for this EALcontains a list of examples of security threats,however, it appears that the individual classifying the event willnot be able to properlyclassify the event without referring to the basis. This EAL should be modified toinclude site specific indications of "security threats" or additional information shouldbe provide to justify this deviation from NESP-007.

~Res ense

EAL8.1.1 has been revised as follows to include specific indications of securitythreats:

"Confirmed security threat directed toward the station as indicated by:

- "Any threat information, the consequence of which would have probableadverse consequences upon vital or non-vital areas of the plant, which isreceived from a reliable source (e.g. Federal, Sate, or local law enforcementagencies, NRC, Corporate Security)."

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28. EAL 8.1.2 Continued hostile intrusion or actin the plant protected area

The corresponding NESP-007 IC to this EAL is HA4, which contains the following asEALS:

1. Intrusion into plant protected area by a hostile force

2. Other security events as determined from (site-specific) safeguardsContingency Plan.

Additional information is needed to justify that "hostile intrusion or act" is consistentwith the NESP-007 guidance for "other security events" and to justify how the SSESEAL is related to the SSES Safeguards Contingency Plan.

~Res ense

EAL 8.1.2 has been revised as follows:

Confirmed hostile intrusion or act in the plant PROTECTED AREA as indicatedby;

~ One or more armed person has breached the Protected Area Barrier.

OR

~ Surreptitious altering of plant equipment or control mechanisms essential tothe safe operation or shutdown of the reactors.

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29. EAL 8.1.3 Confirmed hostileintrusion or actin the plant vital areas~ ~ ~

The corresponding NESP-007 IC to this EAL is HSI, which contains the following asEALS:

1. Intrusion into plant vital area by a hostile force

2. Other security events as determined from (site-specific) safeguardsContingency Plan.

Additional information is needed to justify that "hostile intrusion or act" is consistentwith the NESP-007 guidance for "other security events" and to justify how the SSESEAL is related to the SSES Safeguards Contingency Plan.

~Res ense


Confirmed hostile intrusion or act in the plant VITALAREAS as indicated by:

One or more armed person has breached a vital area of the plant (excludingthe Security Control Center).

OR

~ An explosive device has been found which, if detonated, would have a

negative impact on the safe operation or shutdown of a reactor.

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30. EAL 8.2.2.a Fire likely to affect the redundant trains of one or more plant safetysystems required for the current operating mode

The corresponding NESP-007 is IC HA2, "Fire or Explosion Affecting the Operabilityof plant safety systems required to establish or maintain safe shutdown." One of theEALs for NESP-007, IC HA2, is as follows:

1. The following conditions exist:

a. Fire or explosion in any of the following areas (site specific) and

b. Affected system parameter indications show degraded performance orplant personnel report visible damage to permanent structures orequipment with the specified area

The SSES EAL deviates from the NESP-007 guidance in that no site specific "areas"are listed and in that the SSES EAL specifies that the fire affects redundant trains.

The intent of the NESP-007 guidance for this IC is that only one of the redundanttrains needs to be affected to declare an Alert. The concern is with the magnitude ofthe fire and damage to a single train of a safety system is used as an indication of afire of sufficient magnitude to warrant an Alert to be declared.

This EAL should be modified to include a condition which indicates the "specific areas"affected by the fire or additional information should be provided which justifies notincluding specific areas. In addition this EAL should be modified to change the EALcondition of "Fire likely to affect the redundant trains" to "Fire likely to affect a train."The licensee may propose a different condition for this EAL which willbe indicative ofa fire of sufficient magnitude to warrant an Alert declaration consistent with the basisfor this NESP-007 EAL.

~Res ense

This EAL has been revised as follows in order to provide an indication of a fireof sufficient magnitude to warrant an Alertdeclaration consistent with the basisin NESP-007.

Fire within a plant VITALSTRUCTURE which has affected plant systems required toestablish or maintain cold shutdown.

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EAL 8.2.2.b Report or detection of toxic or flammable gases within a plant vitalstructurein concentration that willbe life threatening to plant personnelor will affect plant systems required to establish or maintain coldshutdown

This EAL deviates from NESP-007, IC HA3, "Release of Toxic or Flammable Gaseswithin a Facility Structure..." in that the SSES EAL specifies plant "Vital Structure"while the NESP-007 specifies "Facility Structure." The basis for the NESP-007 EALstates that "this IC applies to building and areas contiguous to the plant Vital Areas..."This EAL should be modified to include areas contiguous to the plant Vital areas or. ~

additional information should be provided to justify this deviation from NESP-007.

R~es ense

This EAL has been revised to include areas contiguous to the plant vitalstructures. The EAL now states "... within a plant vital structure or within anarea contiguous to a plant vital structure in...".

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32. NESP-007 IC AU1 Any unplanned release of gaseous or liquid radioactivity to theenvironment that exceed two times the radiological technicalspecifications for 60 minutes or longer

SSES did not include EALs corresponding to two of the EALs included under thisNESP-007 IC.

One of these EALs is related to readings on perimeter radiation monitoring systems.SSES states that they do not have telemetered perimeter monitors. A compensatoryindication such as field monitoring readings should be used to provide the readings..which are referred to in the NESP-007 EAL. SSES should include this type of an EALor should provide additional justification for its omission.

The other EAL not included in the SSES submittal is an EAL related to the automaticreal-time dose assessment capability. SSES states that they do not use automaticinitiation of real time dose assessment. A compensatory indication such as manualinitiation of real time dose assessment should be included in their EALs. SSES shouldinclude this type of an EAL or should provide additional justification for not includingthis type of an EAL. This comment also applies to NESP-007, IC AA1.

~Res ense

As stated in NESP-0007 IC AU1 and ICAA 1, the EALs for the parameterradiation monitoring system and real-time dose assessment are only applicablefor those sites having such capability. Susquehanna SES does not have thesecapabilities.

EAL 5.1.1 has been revised into two sections. EAL 5.1.1a is the same as existing EAL 5 ~ 1.1.

EAL 5.1.1.b is as follows:

All Abnormal gaseous release for greater than60 minutes which exceeds 2 timesTechnical Specifications as indicated bysample analysis.

Tech Spec 3.11.2.1

EAL 5.1.2 has been revised into two sections. EAL 5.1.2.a is the same as existing EAL 5.1.2.

EAL 5.1.2.b is as follows:

ALL Abnormal gaseous release for greater than15 minutes which exceeds 200 timesTechnical Specifications for particulates or20 times Technical Specification limits forNoble gas or iodine 131 as indicated bysample analysis.

Tech Spec 3.1 1.2.1

EAL 5.2.2 has been revised to read as follows:

ALL Actual dose at or beyond the EPB. > 10 mrem/whole body> 50 mrem/child thyroid

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33. NESP-007 IC FU1 Any loss or any potential loss of containment

SSES does not include an EAL for loss of containment which corresponds to theNESP-007 IC. The licensee should provide additional information which justifies notincluding an Unusual Event level EAL for the loss or potential loss of containment.

In addition SSES does not include a potential containment loss EAL which correspondsto the NESP-007 EAL "explosive mixture exists." SSES states "for an explosivemixture to exist a LOCA with Fuel Damage must have occurred and that theseconditions are covered by the Drywell Rad Monitor and the Reactor Water Level..thresh olds."

The presence of an explosive mixture in the drywell is a valid indication of the potentialloss of primary containment barrier. Barrier based EALs are not derived from a givenaccident sequence, rather they reference available indications that a barrier ispotentially lost or lost.

"SSES should add this EAL or provide additional justification for this deviation from theNESP-007 guidance.

R~es ense

EAL 3.3.1 has been added to address the issue of including an Unusual Eventlevel EAL for the potential loss of containment. This new EAL is as follows:

1 ~ 2. 3 Containment HydrogenConcentration

>4%

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34. NESP-007 IC FS1 Loss of both Fuel Clad and RCS

SSES did not tie the reactor coolant activity (indication of a loss of Fuel clad) with theDrywelt'ressure (indication of a loss of RCS) as being indicative of a loss of twobarriers and therefore being classified as a Site Area Emergency per the NESP-007guidance. SSES states that they elected not to explicitly tie Reactor Coolant Activityand Drywell Pressure because of the extended time required to obtain a reactor coolantsample. In addition, they stated that this combination is considered unnecessarybecause other indicators cover this event with more timely indications.

Elevated reactor coolant activity is a valid indication of the loss of Fuel Clad Barrier.Barrier-based EALs are not derived from a given accident sequence, rather theyreference available indications that a barrier is potentially lost or lost. It is appropriateand in accordance with NESP-007 to include this barrier-based EAL.

SSES should add this EAL or provide additional justification for this deviation from theNESP-007 guidance.

R~es ense

While there is not an explicit EAL which combines loss of both Fuel Clad andRCS, we believe that this situation is adequately addressed by several otherEALs. These EALs are 1.1.3, 1.2.3.a, 1.2.3.b, 2.1.3.a, 3.1.2 and 3.1.3.Reactor coolant activity is an indication of fuel clad failure. Containment highradiation is also an indication of the loss of both fuel clad and RCS. Reactorwater level and containment pressure also provide indication that the RCS hasbeen lost and thus fuel clad may be lost.

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35. NESP-007 FG1 Loss of fuel clad, potential loss ofRCS, and loss ofcontainment

NESP-007 guidance contains as an initiating condition (FG1) for the General Emergencyclass, the loss any two barriers with the potential loss of a third barrier. SSES did nottie the reactor coolant activity (indication of a loss of Fuel clad) with the RCS leakage(indication of a potential loss of RCS) and with indications for loss of Containment andtherefore does not have in this case an EAL which corresponds to the NESP-007guidance.

SSES states that they elected not to explicitly tie Reactor Coolant Activityand Drywell. ~

Pressure because of the extended time required to obtain a reactor coolant sample.In addition, they stated that this combination is considered unnecessary because otherindicators cover this event with more timely indications.

Elevated reactor coolant activity is a valid indication of the loss of Fuel Clad Barrier.Barrier-based EALs are not derived from a given accident sequence; rather, theyreference available indications that a barrier is potentially lost or lost. It is appropriateand in accordance with NESP-007 to include this barrier based EAL.

SSES should add this EAL or provide additional justification for this deviation from theNESP-007 guidance.

~Res ense

While there is not an explicit EAL which combines the loss of fuel clad with thepotential loss of RCS and with the loss of containment, we believe that thissituation is adequately addressed by several EALs. These EALs are 1.1.4,1.2.4.a, 1.2.4.b, 2.1.4.b and 3.1.4. These EALs use reactor coolant activity,containment high radiation, reactor water level and containment pressure toindicate the loss of all three barriers or the loss of 2 barriers and the potentialloss of the third.

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36. NESP-007 IC HU1 Natural and destructive phenomena affecting the protected area

EAL 7 for this N-ESP-007 IC is other Site-specific Occurrences. Examples given in thebasis fo'r this EAL includes floods. SSES did not include this in their EALs. Internalflooding of the plant should be added as an EAL or additional information justifying notincluding this EAL should be provided.

R~es ense

Flooding of plant areas from external events is not a concern at SusquehannaSES. The plant area is located approximately 4000 feet from the SusquehannaRiver at an elevation of approximately 180 feet above the river..Groundwaterlevels are below the base of the vital structures.

EALs 7.5.1 and 7.5.2 have been added to address flooding of the plant for internalevents. These EALs read as follows:

7.5.1'onfirmed flooding within a plant VITALSTRUCTURE.

7.5.2 Confirmed flooding within a plant VITALSTRUCTURE which has affected plantsystems required to establish or maintain cold shutdown.

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37. NESP-007 IC SS4 Complete loss of function needed to achieve or maintain hotshutdown

SSES d6es not have an EAL which corresponds to NFSP-007 IC SS4. SSES statesthat the NESP-007 EAL is applicable to PWRs and that there is no analogous situationfor a BWR.

The staff agrees that the NF-SP-007 EAL is tailored for PWRs. However the staffbelieves that a similar EALfor BWR should be developed for complete loss of functionneeded to achieve cold shutdown.

R~es ense


A and (B or Cl

A. Loss of all systems needed toachieve COLD SHUTDOWN.See Table 7.3.1.

1. 2. 3 AND

B. Suppression Pool Temperature

OR

C. Suppression Pool Level

200oF

5 12 ft.

Page 38

Download - Pennsylvania EALs,in j · Pennsylvania Power 8 Light Company Two North Ninth Street ~Allentown, PA 18101-1179 ~ 215/774-5151 Robert G. Byram Senior Vice President-Nuclear 215/774-7502

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