NL-15-1898, Enclosure 5: Vogtle Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 4 of 4

Enclosure 5: Vogtle Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 4 of 4

ML16071A174
Person / Time
Site:	Hatch, Vogtle, Farley
Issue date:	03/03/2016
From:	Southern Nuclear Operating Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML16071A108	List: ML16071A110 ML16071A126 ML16071A132 ML16071A136 ML16071A137 ML16071A140 ML16071A144 ML16071A146 ML16071A147 ML16071A149 ML16071A150 ML16071A152 ML16071A153 ML16071A158 ML16071A160 ML16071A166 ML16071A170 ML16071A174 ML16071A176 ML16071A180 ML16071A181 ML16071A184 ML16071A185 ML16071A187 ML16071A188 ML16071A193 ML16071A195 ML16071A198 NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 1 of 4 NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 2 of 4 NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 3 of 4 NL-15-1898, Enclosure 2: EAL Deviations and Differences Maxtrix (Farley, Hatch, and Vogtle) Part 4 of 4 NL-15-1898, Enclosure 3: Hatch EAL Calculations - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 3 NL-15-1898, Enclosure 3: Hatch EAL Calculations - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 2 of 3 NL-15-1898, Enclosure 3: Hatch EAL Calculations - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 3 of 3 NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 6 NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 2 of 6 NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 3 of 6 NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 4 of 6 NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 5 of 6 NL-15-1898, Enclosure 4: EAL Verification and Validation Documents - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 6 of 6 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 2 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 3 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 4 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 5 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 6 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 7 of 8 NL-15-1898, Enclosure 5: Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 8 of 8 NL-15-1898, Enclosure 5: Vogtle Marked-Up EAL Schemes - License Amendment Request for Changes to EAL Schemes to Adopt NEI 99-01 Rev. 6 and to Modify Radiation Monitors at Farley Nuclear Plant. Part 1 of 4 ... further results
References
NL-15-1898
Download: ML16071A174 (44)

Text

HU7 ECL: Notification of Unusual Event Initiating Condition:

Other conditions exist which in the judgment of the emergency director warrant declaration of a (NO)UE.Operating Mode Applicability:

All Emergency Action Levels: (1) Other conditions exist which in the judgment of the emergency director indicate that events are in progr'ess or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist wieh-thai are believed by the emergency director to fall under the emergency classification level description for a NOUE.138 J4,8$YSTEM MALFUNCTION ICS/EALS GENERAL SITE AREA EMREC EEGNYALERT UNUSUAL EVENT SG1 Prolonged loss of SSI Loss ofall offsite SAI Loss of all but one SUI Loss of all offsite all offsite and all onsite and all onsite AC power AC power source to AC power capability to AC power to emergency to emergency buses for emergency buses for 15 emergency buses for 15 buses. 15 minutes or longer. minutes or longer. minutes or longer.Op. Modes: Power Op. Modes: Power Op. Modes: Power Op. Modes: Power Operation, Startup, Hot Operation, Startup, Hot Operation, Startup, Hot Operation, Startup, Hot Standby, Hot Shutdown Standby, Hot Shutdown Standby, Hot Shutdown Standby, Hot Shutdown SA2 UNPLANNED SU2 UNPLANNED loss of Control Room loss of Control Room indications for 15 indications for 15 minutes or longer with a minutes or longer.significant transient in Op. Modes: Power progress.

Operation, Startup, Hot Op. Modes: Power Standby, Hot Shutdown Operation, Startup, Hot_________________Standby, Hot Shutdown SU3 Reactor coolant activity greater than Technical Specification allowable limits.Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU4 RCS leakage for 15 minutes or longer.Op. Modes: Power Operation, Startup, Hot________________Standby, Hot Shutdown SS5 Inability to SA5 Automatic or SU5 Automatic or shutdown the reactor manual trip fails to manual trip fails to causing a challenge to shutdown the reactor, and shutdown the reactor.core cooling or RCS heat subsequent manual Op. Modes: Power removal, actions taken at the Operation Op. Modes: Power reactor control consoles Operation are not successful in shutting down the reactor.Op. Modes: Power__________________

_________________Operation

___________

139 GENERAL SITE AREA EMREC EEGNYALERT UNUSUAL EVENT SU6 Loss of all onsite or offsite communications capabilities.

Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU7 Failure to isolate containment or loss of containment pressure control.-[P-WR{

Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SG8 Loss of all AC SS8 Loss of all Vital and Vital DC power DC power for 15 minutes sources for 15 minutes or or longer.longer. Op. Modes: Power Op. Modes: Power Operation, Startup, Hot Operation, Startup, Hot Standby, Hot Shutdown Standby, Hot Shutdown SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 140 SG1 ECL: General Emergency Initiating Condition:

Prolonged loss of all offsite and all onsite AC power to emergency buses.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director sheui4dxill declare the General Emergency promptly upon determining that (sie-pee hours-) has been exceeded, or will likely be exceeded.(1) a. Loss of ALL offsite and ALL onsite AC power to specific emerge.ncy AND b. EITHER of the following:

• Restoration of at least one AC emergency bus in less than speeifel hours:) is not likely.* ...... .(S.it:* indic.-tizn cfon'Basis: This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers.

In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should-wilI require declaration of a General Emergency prior to meeting the thresholds for IC FGI. This will allow additional time for implementation of offsite protective actions.Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.The estimate for restoring at least one emergency bus she~*1d-wvill be based on a realistic appraisal of the situation.

Mitigation actions with a low probability of success eheul4-witl not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.i~¶19~5TtntQmWi~

141 that._ cupp~ly pc;wzr tc. thc elcctrical di.c.t:r:ib:u.ti.c.

cyctem that po-ve.-c SAFETY SYSTEMSE.

Th.c.rc.ic yial I, zmzrg.ency but; pr train: of S.AFETY SYSTEMS.....iz ... H+ ...baetkut coping an"aly:°ic prform:'ed in accrdanzz+

wiqth 10 CFR. 50.63 rand PRegulato.ry G'uic, 1. 155, Stationx B!azc!.o:.

Site spceific indication of an' inability3 to adequetey rez' heat from thc core:[PUll1 !nocrt sire cpoeiic v'alues for an incor'e/carc exit thz,.rmozouplz tcmeraturc

°andor irPotr-men~tatin that cannot mneasure do--n to approoiroaely thc middle ofr aeti;'e f"uel ohou.:d u+z the !cv.zzt onr scal rcoding that ic not "ab-ve the top of acti'r fu".cl. If the !l',- et en sLereadin~g ic abov... the top o.f activ'e fuel, +the a........

r.at.. ccc .,vc , ...u. chou.ld not be. incud,..F,.r planto th.+ hav... implemented Westinghouse Orwnere Group tEm.........

pReoonce Gui.:d.eline cntcr the par-t-ro uted nteCr e C....... ling... Re Path....ECL Aoionn~ent Attributeo:

3. l.1.B 142 SG8 ECL: General Emergency Initiating Condition:

Loss of all AC and vital DC power sources for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director sheuildvd ii declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. Loss of ALL oflfsite and ALL onsite AC power to BO11 )All~-': site specific

......uses) for 15 minutes or longer.AND b. Indicated voltage is less than (ies if b ;ztgvauAn ALL for 15 minutes or longer.Basis: This IC addresses a concurrent and prolonged loss of both AC and vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.Fifteen minutes --az selected ais th.e threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.'Dnepe'e-rNetes:

The eitec speeifie em'.ergency buses" are the buses fed by' oft'tte or em..ergncyc3 AC peWer scures~The. 'zite ......i ;b, c v..e tage "'alu'" eh-"uld b.^ b.aced en the hu:s ;vltage neccc=a.-y for adequate e~peratiea af SAFETY SYSTEM. equipm-e~nt.

This ;veltage ;'aluae seheu!d the cd.The v-altue .,i .........

nea the minimum. : ....a.. sel..t. when. barte- c i.- .....ee the. cell.. veltage1..

is apprax:matel 1.7 'J.... t e~t per e]cU. For a" 5° string ba..e.. th.e minimum= voltage is appraximately 1.81 Vclts per ol....143 Th,. -i;t. zp~cific Vita.l DC,-, ...... a thc, DC b. .... tha ........ ; cn i -;,.n... an ...... I fcr SAFETY $YSTEMS.LLL AZ~IgnmCflt Attrltutzz: .j. I.1.Ii 144 SSl ECL: Site Area Emergency Initiating Condition:

Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director i declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(I) Loss of ALL oflfsite and ALL onsite AC power to (site .....ci.... e ...: mergen.cy' 2)AAQ2 1()B +for 15 minutes or longer.Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.Fifteen minutes ;v'as ais th.. threshold to exclude transient or momentary power losses.Escalation of the emergency classification level weald-beauscs ICs RGI, FGI or SG1.minimu n.. umbr of cpernting+

generators e sr f... r. that source........

pr'vi adequate power to+i.e...to5%cpct g,+ne_ t..r ;+.i......t.

fced I+ e.+ mergeny bus, te[Lan ai seto utseiyta ehgnrtr o h creacceaig The. site spciic+,, emergencyl~

buses are the buse fcd by offsite or emergency AC po +.er sour that suplyl, power+ to the+ electrical ditributio., system. that, l-,wcrs+ SAFETY

.Theretyicll Il emegenc train +IVI of SAFETY SYSTEM UV =UII Th.[A an/o .. as....s. sectio may. .pe.ify u..e -f a n..n aet r.. lated. po, sou.+:-rc++-':e+

prvided that iperotion of,. this source ., 1.++ is contr.lle..in acodac wit abnora or. emergenc.

oprai 145:; + + + + ;++++++;++

+++ + + + + + + 1 +

crcccdur~&.

dc. Iicnt that fla~c a t~rzccduraiIzcd ca~atiIitv tc ~uDch dttzItc AC ~cwcr tc ~e#I~ncd unit via a crz~ tic t~lici that thc pI~r.c~ crc a zn*a ni a cci thi n.. * ,th ac tic ~tmtcgy mcct thc rc~1uircmcnt~

cf 10~R-~O~63~ECL Mcignmcnt Attributc~:

3.1.3.8 146 SS5 ECL: Site Area Emergency Initiating Condition:

Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal.Operating Mode Applicability:

Power Operation Emergency Action Levels: (1) a. An automatic or manual trip did not shutdown the reactor.AND b. All manual actions to shutdown the reactor have been unsuccessful.

AND c. E ITHER of the following conditions exist:* specific indication of an to adeouatelv remove heat from the RC~)Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs.

This is appropriate i~-thabccausc the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Escalation of the emergency classification level ;weuid be- 'auses IC RGI or FGI.po--er le-vel that is: !ess, than or equal to the reactor po--er la-vel which defines the low:.er be'und ot" 147 Applicab~iit.

Far example, if the rcaatcr i.. c.n.idered ta be .shu.tda-,n at 3% an~d Pa-,-er Oeraticno, strt .t....thn.h......a....picb i n Start.p...d.

Dev..... r may.... in'lude site specific EOP crite-ia indic~ativez sfa successful

....ct.r hut.da....

in an EA tat cmo ... the B. is beth (e.g., a reactar pewver leel..." Site specific indication af an inability ta adequately rema-ve heat from the care: tPI4R

.. ite. s..'pecific v'alues fcr an inccre/care therm:cocuple temperature and.'cr rFor vessel whate lavel that drives Wentryinthcore Ow alner procedure (rchrv exit thennacouple" tem.raure great:er th.ana::ae'

-,20an.'ar ater vl ha!o ,spn ds2A ' tcePW apxitely, thcen Pmidl:eof .activer fu~el. Pat ihratrv eee EC L A,'ribuates:

3.1!.3.B 148 SS8 ECL: Site Area Emergency Initiating Condition:

Loss of all vital DC power for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director ehe~ldwilI declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Indicated vote isls hn(ieseii u ctg au LL ~I 12- ) forl15 minutes or longer.Basis: This IC addresses a loss of vital DC power whieh-that compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.Fiftee minues ..a...l.t....

ais the threshold to exclude transient or momentary power losses.Escalation of the emergency classification level -,.'ild be-v4auses ICs RG1, FGI or SG8..:c :c ;pe::s u:.':=o;au aacne a; :: une tha:n'muml:;

.:vs ~ge ne.......

..... fe..r' operation cf SAFETY SYSTEM equipment.

Thi's .. vcltage '.alue shoeuld inearpe-.t n+ .... margin ef at least 15 minutes of operation b.efere. ...h.e enset of inability to operate thee lads. This voltage is usually near th-e minimum voltage selctd he b.attery ......

• i*Th.. typi.cal -'lue for.an.ent.r..b.t... .et 105 .DC .Fcr 60 u cll trin eof batteries, the,. c,.ll .,cl tge .. is appr--imat-l-

.7;.. 5 Volts per cell. a" 50 string batter. set, the...control capabilities for SAFETY SYSTEMS.ECL A s r...........

ribut......

.l... ' B 149 SA1 ECL: Alert Initiating Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director sheu~dwill declare the Alert promptly upon determining that 15I minutes has been exceeded, or will likely be exceeded.(1) a. AC power capability to spezitic cmzrgenqcy bu:;e;)lOTlI I ) I 2 is reduced to a single power source for 15 minue r longer,...

AND b. Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.7*~: Im~pcy~T~wm5 y~s~Th~w~nz Basis: SA\FVIY SYSI 'M: A systcen required lb)r sal'e plant op~ration, cooling down thc plant and/or placing it in the cold shutdown condiiion, including the [I('(S. 'lhcse are tvpficaill systems classified as safety-rclatcd.

This IC describes a significant degradation of offsite and onsite AC power sources (see lFable S I abo'.e) here any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

This IC provides an escalation path from IC SU1.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

150

• A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.

• A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.Fifteen minutes sel'ected as a is the threshold to exclude transient or momentary losses of power.Escalation of the emergency classification level w:ecu!d-be-*iauses IC SS1.minimum. number of

+ generators n..e..a..

for tat our..... to pro.id. required.

po ..... t an A~Ca!, emerg.n.y.bu..

Far xmlifabcu pwer trurac iscomArrsc of two enerator Deveclopers ashuld m,,odily the bulleted ersamples pros'idedl in the basis seetirn, above, as needed Incoming.....

powr .ines) c ompis thr sepa...te.p..e...our.es.

Independence may be determ~ined fro a... reie of.. site UFA, SP or.. , los of electrical pow'er The. EAL

... Bai s ..ti .may speci' u...of a non.. sa.... related pow;er source prcvided At multi unit stations, the EALs may credit compensatory measu,-es that are proceduralized and'.sing generators,..

+- other:_ pe soures ,escibe gs anr al or emergen- cy~ op resting .........

s,. et. Plnt thato .++ hav av procedura ied capability t supply..+

offse AC power to an ECL Assignment Aftrbutes:

3. l.2.B 151 SA2 ECL: Alert Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: SNote: The emergency director sheo4will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the control room for 15 minutes or longer.Reactor Power RCS Level RCS Pressure In-Core/Core Exit Temperature Wide Ranse_ Levels in at least (Ite ... .........:r... steam generators Steam Generator Main or Auxiliary er Emerge+acy-Feed Water Flow AND b. ANY of the following transient events in progress.* Automatic or manual runback greater than 25% thermal reactor power* Electrical load rejection greater than 25% full electrical load* Reactor trip*

actuation Basis: UNIPIANN I-D: A parameter change or an event thai is not I) the result of an intended evolution or 2) an expeeted plant response to a transient, The cause of the parameter change or event may, be known or This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the control room. During this condition, the margin to a potential fission product barrier challenge is reduced. It t-has-represents a potential substantial degradation in the level of plant safety-of-th As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the control room. This situation would require a loss of all of the control room sources for the given parameter(s).

For example, the reactor power level cannot be determined from any analog, digital and recorder source within the control room.152 An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.

The event wotl-beis reported if it significantly impaired the capability to perform emergency assessments.

In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency pian implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.

In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control hoard, the SPDS or the plant computer, theni the availability of other parameter values may be compromised as well.Fifteen minutes -waz soeacced ae a.ir, the threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level ICs FSl1 or IC RSI1.cr ~Noee: Iii-4he-PWPD carameter ILt eelumn. the ~eite er.ecitie number eheuld reflect the t~inimum num .. ....r .f.am g n...... .. .. .... ..x.....a ...f r .l ..t ......n. ... .... ew........

.... ... ..m.y...epecif whether the ........lu .......d.b ...id .rage ........ .. ran ..r ......th ........g u:Fn th..antr.' for P.CS Level.The. number,

....ie anda laycu.:t ef Centre! R.e.am- ..nd.ca..en.,:':*^-

and the range of pez.eible peio available.....f..r emrgen.y c!oeeifieatio~n a~eczmcnts.

if."- zpecific porer..n:g ef indi!catiene By° fo.u.ing ..n. the avallaiit of the, spcie pa... amet..r...alu..., in.t.a...f.th............th..

vaue,m the. re...i. e an... ac.......ate the. wid. ...i.. ty' of indicatio~ns in nuclear pow;er A loss ef plant annu'.nciators will be for reportability in .accerdance wi.th 1 0 CFR 50722 capatuitit to p~ertor,'m cemerncn.

areseesments.

te..om.pe.salery icr a" less or 153 annunciation can be reodily implemented and may include increased monitoring of main control boardo and more frequent plant roundc by mm licenced operatorc.

Their alerting function notwithotanding, annunciatoro do not provide the pan meter values or specific compOncnt status infcrmation used to operate the plant. or process through AOPs or EOPo. Based on these considerations, a losa of ~~unciation is considered

~o be adequately addressed by repottability criteria, and therefore net included in this IC and EAL.With respect to establishing event severity, the response to a Icos of radiation monitoring data (e.g., process or effluent monitor values) is eonsider~d to be adequately bounded by the requirement of 10 CFR 50.72 (and associated guidanee in NUREG 1022). The reporting of this event will ensure adequate plant staff and NRC a ~ and drive the establishment of appropriate ecmpensatcry measuree and corrective actions. In addition, a loss of radiation monitoring data, by itself, is not a precursor to a more significant event.Personnel at sites that have a Failure Modes and Effeets Analysis (FMEA) included within the design basis of a digital l&C system should consider the FMEA information when developing their site specific EALs.Due to changes in the configurations of SAFETY SYSTEMS, including associated instrumentation and indications, during the cold shutdown, refueling, and defucled modes, no analogous IC is included fcr these modes of operation.

ECL Assianment Attributes:

3.1.2.13 154 SA5 ECL: Alert Initiating Condition:

Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.Operating Mode Applicability:

Power Operation Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Emergency Action Levels: (1) a. An automatic or manual trip did not shutdown the reactor.AND b. Manual actions taken at the reactor control consoles are not successful in shutting down the reactor.Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful.

This condition represents an actual or potential substantial degradation of the level of plant safety-ef-the An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.A manual action at the reactor control consoles is any operator action, or set of actions, ie that causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies.

If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers).

Actions taken at back-panels or other locations within the control room, or any location outside the control room, are not considered to be "at the reactor control consoles".

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS5. Depending upon plant responses and symptoms, escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either IC SS5 or FS 1, an Alert declaration is appropriate for this event.155 It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.the pewer !t-'e] at* wh*ich the reactor ic cneid!ered

eh-tdo'wn.

A PWPR ":.ith a ohu.:tdo,:,:

reactoer Opwrlcl" that+ i,,tc- ++ -.,o/_ +th_ or equal to th raco pp wrl!lwi! eic h lwrbuda..wer.Oper.tion.(M .de .)..ill.nee..t..in.l..

Starup...

d 2 ......in.. O ..r..in .Mode,, term (e +.:.g.. man onr A boardo)156 SA9 ECL: Alert Initiating Condition:

Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1) a. The occurrence of ANY of the following hazardous events:* Seismic event (earthquake)

• Internal or external flooding event* High winds or tornado strike* FIRE* EXPLOSION* Other events with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the following:

• Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.Basis: I:IRE: (/ontbustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation ot flame is prefe~rred but is NOTI required if large quantities ol smoke anod heat are- observed.E;XPLOSION:

A rapid, violent and catatstrophic fatilure of a piece of equipmenlt due to combustion, chemical reaction or overpressuri/ation.

A release of steam (from high energy lines or components) or an electrical component tailure (caused by short circuits, grounding, arcing.etc.) should not automatically be considered an explosion.

Such events may require a post-event inspection to determine if' the attributes of an explosion are present.SA:E [IV SYS [EM: A system required (or safe plant operation, cooling down the plant and/or placing it in the cold shutdown condlition, incltuding the FCC'S. These arc typically systems classified as saetyt-related.

VISIBLE DAMAGE: Damage to a component or structure that is readily observable ,avithout measurements, testing, or analy sis. t'he visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affecoted component or structureDamage-t'o a ...........

c :-tinig SAFETYt' SYSTEMS v'zual imac ........ cau ., a.b abou th..a zperabi!ity, af SAFETY SYSTEMSa within: .t. ........ u .... Eapz ,i pa..tial ......a, 157

~cii'a~e e3t er a ..........

cngu--d in t+'~~ ..... nzt includ..e ..... su......t,. nz ....tz zr -- --caizc nr.tc,+:.

splling,h

..... I.,fcrmnt.n

...

Ti Lin--c This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of plant safety[hc first threshold tor EAL l.b4-I addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.

The indications of degraded performance shudwilt be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.fh.c second threshold for EAL I b- addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.

Operators will make this determination based on....~i~..... :hll available event and damage report information.

This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.Escalation of the emergency classification level w;eu!d-be-v4auses IC FS 1 or RS I.Nuclar p......r plant SAFETY SYSTEMSc are eomprit;ed

f. t... or mere c~eprate .an redundant.

trano of equipment in .......ne with... ote criteri:.ECL Accie-nment Attribu+teo:

3.1.2.13 158 SUl ECL: Notification of Unusual Event Initiating Condition:

Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director sheutidwill declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss of ALL offsite AC power capability to (site spezific emergenczy buzeg)Bl~J'l___I

~ ~for 15 minutes or longer.Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources (sce table$2 c) renders the plant more vulnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of plant safety- fhe...eplant+.

For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.Fiftee minues ..........

d.n ais the threshold to exclude transient or momentary losses of offsite power.Escalation of the emergency classification level we**l-b-..,se

.C.S.l.ic typically.....

1 --* emergenc .......u.. per train a~f SAFETY SYSTEMS.can be implemented "w;ithin 15 mi:n....ute Canidr ...... ac p....................,:

159

... .t .....LJLA~xgnmcntAttntutc~:

J2.I.A 160 SU2 ECL: Notification of Unusual Event Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the control room for 15 minutes or longer.Reactor Power RCS Level RCS Pressure In-Core/Core Exit Temperature Wide Range Levels in at least (1tte vp.-ifx. ... ...r,umb cin steam generators Steam Generator Main or Auxiliary ef EmrecyFe Water Flow Basis:.iNPLANNFl[):

A paramecter change or an event is not I) the resutll of an intended exolution or 2) an expected platnt response to a transient.

The cause of the parameter ehange or event ma'.be knlox\ n or ulnknloVxFn This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the control room. This condition is a precursor to a more significant event and represents a potential degradation in the level of plant As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the control room. This situation would require a loss ofall of the control room sources for the given parameter(s).

For example, the reactor power level cannot be determined from any analog, digital and recorder source within the control room.An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.

The event welbis reported if it significantly impairsed-the capability to perform emergency assessments~.--, particularly thoseln e,,m.rgc ......assessmeents necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to 161 determine one or more of these parameters from within the control room is considered to be more significant than simply a reportable condition.

In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, then the availability of other parameter values may be compromised as well.Fifteen minutes wac seleeted ac ais the threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level ";;ukl-be-v4auses IC SA2.In the PWR param.eter

!izt cclmn t...he ..it. z* pecif number" should refec t..he minimum.number of steam generaterc neeesza&'

fcr plant cooldewn.-

and shutda-;n.

This may. noce Dev*elepers may specify either pressurizer ar reactor .,essel..1 l...l in th.e- ÷1-,WR. paI'rame.

,. ....lum .. ..entry for P.CS Level.fallure challenge the abiit c f an eperater t .... ac atl d..termin...

wihi the.tim.period av'allable for emergency classitication assessments, if a specitic percentage af indications have been lost. Th.e approach used in this EAL facilitates prompt eand accurate emergeney classifloatian assessments by f ..u..ing on the indiatiensfor a .elect. sub... t o fa B"fcusaingi a th e afac'm,~ ef. they sp~eci..iedt paompenster' vales,-insta ofr th !saus of the vnaluensnuedtaoert the EA rcgnze ndo aproedsts t~hecg AOPe arietyPaf iniation in nuhearpse plont CenrolRo niationsatys opcs -:.d andi cos'ereed ma: be anal"og. *er diaddesafet y relcated or net........m....

p..a.... ltenat..idi.dua mete .a.u. :ar.a. c eptr adequp ipate,) ate. (rnd te asentci, a-. (n sated guianeinNUE 122,and c r "nre if it, significantl impair.. 162 thcir zitc EALs.oar~lc~us IC is inclu-ad fcr modes zf cvrat.icn., ECL Aai~nmznt Atthbutc~:

3.1.1 .A 163 SU3 ECL: Notification of Unusual Event Initiating Condition:

Reactor coolant activity greater than Technical Specification allowable limits.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2)(2) RCS coolant sample acetitil value iladicatinlg

[hi clad degradation greater than Technical Specihication allov\,able limibs as indicated bAYof the tbllowing:

~rA~R. 6&A~ItY limit apecitica in I ccnnieal ~pecitieatian~.

Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications.

This condition is a precursor to a more significant event and represents a potential degradation of the level of plant safety-ef Escalation of the emergency classification level weuld -be- auc ...F.orte.ecgito Category R ICs.-* An installed radiati.en monitor on the letdo~e.ve:;tern or air ejecter.The manitar reading "valu:ec should cofe'esp-.nd te. an P.CS nect-vity level apprevimately at....ch.nical ... Specifi atAn; Talla ab le

..lk. .. .. 31'164 ECL ,ccignmcnt At",ributc;:

3.1.1! .A an'd 3.1!. !.B 165 SU4 ECL: Notification of Unusual Event Initiating Condition:

RCS leakage for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2 or 3)Note: The emergency director shoutld~x ill declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(I) RCS unidentified or pressure boundary leakage greater than ;'aluz)1! for 15 minutes or longer.(2) RCS identified leakage greater than sp..ifi" value)2$ 15 minutes or longer. [(3) Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.Basis: This IC addresses RCS leakage whieh-ffayihat could be a precursor to a more significant event.In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of phlan safety of the plat.EAL #1 and EAL #2 are focused on a loss of mass from the RCS due to "unidentified leakage","pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications).

EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus-apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage in a PWR) or a location outside of containment.

The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications.

Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation).

EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification.

Fer' PWPs,-a~n emergency classification weo.1drequired if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.Escalation of the emergency classification level wc, uIdbe-v4auscs ICs of Recognition Category R or F.166 9~op~Ne#eM EAL #1 Far tha ~ita ~pccific Icak rata ;aluc, antar tha highar af 10 gprn ar tha ;'aluc cpccifizd in tha ~itc~ Tcchni~a1 Spccificaticnz for this ~ af1z~o~c.EAL 1t2 Far tha ~ita ~pacific Icak ra~ ;'aluc, antar tha highar af 25 gprn ar tha valuc ~pcaifizd~n tha ~itc'~ Tachnizal Spccificatiznz for thiz ty~c of Icakaga.For zitz~ that ha~'z Tczhnizal Spacificatiana that Ja nat ~pczify a Icakaga typa fcr ztaam gancmtar tuba Icakaga, dc~'ctapcr~

zhauld inaluda an EAL far tuba Icakagz grcatcr than 25 gpm for 15 minutca or longer.ECL Az~ignmcnt Attributa~:

3.1.I.A 167 SU5 ECL: Notification of Unusual Event Initiating Condition:

Automatic or manual trip fails to shutdown the reactor.Operating Mode Applicability:

Power Operation Emergency Action Levels: (1 or 2)Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

(1) a. An automatic trip did not shutdown the reactor.AND b. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.(2) a. A manual trip did not shutdown the reactor.AND b. EITHER of the following:

• A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.* A subsequent automatic trip is successful in shutting down the reactor.Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of plant safety-ef the-plant.

Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.168 A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip).This action does not include manually driving in control rods or implementation of boron injection strategies.

Actions taken at back-panels or other locations within the control room, or any location outside the control room, are not considered to be "at the reactor control consoles".

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA5. Depending upon the plant response, escalation is also possible via IC FA1I. Absent the plant conditions needed to meet either IC SA5 or FA I, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.* If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and ,heo1d-u ili be evaluated.

• If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

ThL!ICisa pplicabl in.an ...d ..in ..ich .th ...tu .l.ra ...r ......l......ul.........h Li..L .~.ssIgnment

,'uirllsuses:

i.I.z.r~.169 SU6 ECL: Notification of Unusual Event Initiating Condition:

Loss of all onsite or offsite communications capabilities.

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (I or 2 or 3)(1) Loss of ALL of the following onsite communication methods: In planft telephones Public address system (2) Loss of ALL of the following ORO communications methods:[INN (F-mcrgcnc)

Notification Netwo(rk)[(ormmercia!

phones (3) Loss of ALL of the following NRC communications methods: I NS un FederaI I clecoinmunical iob S\ siefm IFfS)[Commercial phonties Basis: This IC addresses a significant loss of on-site or offsite communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.This IC ill be assessed only when extraordinary means are being to make communications possible (e.g., use of non-plant, privately owned equipment; relaying of on-site information via individuals or multiple radio transmission points: individuals being sent to offsite locations,-ete.).

EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.

EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration.

The OROs referred to here are the states of Georgia and South ('arolina; Burke ('ountx. Gecorgia:

Aiken ('ount\. South Carolina:

Barnx.elI and Allendale.

South C arolina: and the Sasannah River Site in South Carolina( ~~cprNt~.EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

170 telephonec, page party cyscteme, radic., tct.). Thic licting should in~clude inctalkxl pla~nt methodc u--ced tao perfrm initial emergency to 0R03 ac deccribe'd in t.he te!ephone linac, commercial telephene lirn.s. radioc:, eatellite te.!ephesnec and in...... uaced cemmuni+. .caticac...

+ t tc ehn4....clo 3+Emergecyc Plan, liting should ...........

t..led plant.equipment and cempenents., and..... -. iteme7+'C' ewe an mainaine ind...dua....The......................i..l.y th dedicated..

LUL' Accien....t Attr+ibut,+:

.t 171 SU7 ECL: Notification of Unusual Event Initiating Condition:

Failure to isolate containment or loss of containment pressure control.fP-uJ Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2)(1) a. Failure of containment to isolate when required by an actuation signal.AND b. ALL required penetrations are not closed within 15 minutes of the actuation signal.(2) a. Containment pressure greater than .. pzzific

............

ImW[AND b. Less than 4 ('1I NI In cnoolcr andc one full train of (" NI I

... .... ee.....:e.... r is operating per design for 15 minutes or longer.Basis: This IC addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of plant safety-ef For EAL #1, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure);

a failure resulting from testing or maintenance does not warrant classification.

The determination of containment and penetration status -isolated or not isolated -se*dv ill be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.EAL #2 addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible.

The inability to start the required equipment indicates that containment heat removalidepressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner.This event weouI4d-xill escalate to a Site Area Emergency in accordance with IC FS 1 if there e is a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.172 FnALth #2~ zp!ii. pe a-u[!r"~ to "z tha. E:tuatxti.

ECL .AsIznment Attr:'5utz:

3. !. 1.A.......c.tp z r c rI o , z(...tary Pov.~r Kza~tcr (EFK) ~ie~ign.173 APPENDIX A -ACRONYMS AND ABBREVIATIONS AC ......................................................................

Alternating Current AOP......................................................................

Abnormal Operating Procedure PRM .......................................................................

~. Aragsz Povwr Ra:nge N1,.eter ATWS............................................................

Anticipated Transient Without Scram IL .....................................................................................

....Bin c lz an.... .....lT................................................................ nta~............B'

................................................................................................

B W R ..........................................................

ioling W aktr Reactcr CDE .........................................................................

Committed Dose Equivalent CFR .........................................................................

Code of Federal Regulations CTMT/CNMT

..........

....................................................................

Containment CSF ..............................................................................

Critical Safety Function CSFST..............................................................

Critical Safety Function Status Tree DBA...............................................................................

Design Basis Accident DC ..........................................................................................

Direct Current EAL.............................................................................

Emergency Action Level ECCS..................................................................

Emergency Core Cooling System ECL.....................................................................

Emergency Classification Level I ..........................

.....................

I ncngc nc.x N or i Ilecal ion Nct ~ ork I ]N'...........................................................................I.

Imcu~cncv Nolilicalion Systemn EOF ......................................................................

Emergency Operations Facility EOP ....................................................................

Emergency Operating Procedure EPA...................................................................

Environmental Protection Agency PGIP..................................................................mr ny Plargnv lPrled'-ent Prax.durin PRP................................................................EEltignnry PIov!crme"t!.

Pr-..c~dz-ER..................................................................................Em rc.................

P~d 4Re........

..........

I ..........................................

Aviaution," AP minise r ationk, FEqdMeA..........................................................Fede dral EmrgnciaiinAgementlageoncS.......................................................................

Fc dcralII I Bo reau onfca ionvs t cinai FEMA .......................................

..................................

FeraEmgnyMngmntiAgeniy II,-..............................................................................................Fcl

..................

R..................................................

Final.......a....t

...............

R Fz GE .....................................................................................

General Emergency A-!

................................................................................

it.....................

Li i TL.................................................................

P!........................................................................

Hg ol,: r......H.................................................................

Fl............................i

...................

RC SIC.................................................................................

r.Initiatin

!Conditio ID(().....................................................................dqmtrsO aIonsid Diame te(r(................................

.................................................................

PEEF........

.. .lnhda Plan Exa~minatian af External ELent, (Generic Lectcr 88 20 ISFSI ......................................................

Independent Spent Fuel Storage Installation Keff ............................................................

Effective Neutron Multiplication Factor.................................

..............................................................

1.t.....................................................................................................

Liden OC ...............................................................................Lss f",,°,tAcd.Nt CR .......................................................................................

Main Ccntral Lcp mR..mme, RM...............................................milli-Roengen.Equival nt a NEIX.............................................................................Mi Nutear Energyia Inaltue NR ...................................................................................

mmisio NQAS ...........................................................

MNrhAmrcniopaeDfnstea Limmnd (NO)U m ........m, ..........

re............mREM...................m(Notificatgen Eq Uuival EentMa I:-AR" ...........................................................

NtcrNaIg::crt"rdRscrc cr l NBEI...........................................................................Nulearatn ai EnergIntthutke NPP ...............................................................................

Nulear PoweroPledAnta NRCM...................................................

NufstDs ClarlRegla Co...........xMmisinua OO..................................................................Off-ite.Rspons.Orgaizatin PA......................................................................................Nlartm uPplytected re.....N..............................................................Noti.....cation.....O......nusual...

Evn PG......................................................................................Gidelne

..........

C.................................................................Owner......Controlled......

Are PA ...............................

....................................

Pccs!f,- ...Proad Aeasc A-2 RAIPSA ....................

Prcbabil!ztic Rbak A~ocmz.'n..t

/ Prabablti.'tc, Safety ,*a~o~.anwn PWR ..........................................................................

Pressurized Water Reactor PSIG......................................................................

Pounds per Square Inch Gauge R...................................................................................................

Roentgen dCC............................................................................

Reatar Care i rlti! Cnzaln R(D1 .......................................................................

Reactor (oolant lrain Tanik RCS ..............................................................................

Reactor Coolant System Remn, remn, REM..............................................................

Roentgen Equivalent Man RPS ...........................................................................

Reactor Protection System RPV ..............................................................................

Reactor Pressure Vessel RVLIS ..................................................

Reactor Vessel Level Instrumentation System V/CU .......................................................

Rearctar W.atear Cleanup SA ...........................................................................

Site Area ntcmrgecncx AR ......................................................................................

Saet ...... AS................................................................................

S .SDAteat.ian Syskate SC. ........................................................................................

South Carolina SCBA .............................................................

Self-Contained Breathing Apparatus SG ........................................................................................

Steam Generator SI ..........................................................................................

Safety Injection leJ \ ..................................................................

Sfr*!:a'.atciun Jet CAir I Sjcctor SN ..........................................................................

Southern Nuclear Compan?SPDS ..................................................................

Safety Parameter Display System TD...................................................................................quialen TOAF.................................................................................Seniprof Active Fuelr T F..............................................................................Top....of...Active....

Fue SC ..............................................................................

Technical Suppert Conter VI)C ................................................................................

Volts lirect Current\I( I. ...................................................................

Vogtle F Iccritt ( icncrat in a Plant V()IP ...................................................

oice O~cr lnternet Protoeol W tl I ..........................................................

A asta I loldlup lank OG .........................................................................

Wcatinghzuec Ow-.ncr5 Group A-3 A-2 APPENDIX B -DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents.

General Emergency:

Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.Site Area Emergency:

Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; I) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.Alert: Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION.Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.Notification of Unusual Event Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.The following are key terms necessary for overall understanding the NEI 99-01 emergency classification scheme.Emergency Action Level (EAL): A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level.Emergency Classification Level (ECL): One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (I) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in -dtcsccnding order of severity, are: General Emergency (GE)Site Area Emergency (SAE)Alert Notification of Unusual Event (NOUE)Fission Product Barrier Threshold:

A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.B-I Initiating Condition (IC): An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.

Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document.

The definitions of these terms are provided below.CONFINEMENT BOUNDARY:

I1 he barricrls) between areas containing radioaeti\xc substances and th.. enviorii'n.(l"zr a ......i d*:' i " itin fre thit5 tc..rr.) ,a .... r Na.. : ba.ri:r.,)

bu c .........

fu.c, an.d tz ..... n....t zn t... fue"l i; pracc~cd fcr dry, CONTAINMENT CLOSURE: l'cr Operating Procedure 14210-I 2. C ontainment Ituilding P'cuclt iaons Veritication

-Retie in ,~ f a itx~ .: ....... : d. .in.i:, an for.u: thi tc n .) .n .....Noe= The prczedur=ly defined ar taken ta. zae-ra zant.ainmcnt (7primary, r EXPLOSION:

A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.

A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Such events may require a post-event inspection to determine if the attributes of an explosion are present.FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

Natc.e Thic t.... : .....~bl ,:t., P,. an... .FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.HOSTILE ACTION: An act toward a nuclear poWvCr plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (0(A }).HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

B-2 N cI 99 01 29:z !2 IMMINENT:

The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.OWNER CONTROLLED AREA (OCA): The site property os'ncd hx or under the control of V[IG' .eort... * ... a cte .cpccfic def.n.tien fr. .: thc .nn.) Df-cvzlopr ,Nate -,Thic perimeter to... th.. plant Patcte Ar...a perim.ter ( .... g ... a* cite th a large OC A w:here some deap..... , .... d .on. i ... : u ...i:n g. the b.undary, defned by the Rec.tricted a~r Secured O-w.ner PROJECTILE:

An object directed toward an NPP that could cause concemn for its continued operability, reliability, or personnel safety.PROTECTED AREA (PA): [he area that eneotpa sses, all controlled thinhh the sceurity protected area fcner l nc cr. cite.....

cpeii .................................

p rN .....h ...t..

t...n t. me.... the ......a under .: .............

~menitacring an'd centrel, and a.rmed REFUELING PATHWAY: l his includes the reactor refuel cavil',lhe fuel lran~slkr canal, and the Sl'pnt tue pool. canals and pools through x Ihich irradiated tuc I ma\ he moved, hot n/ot ineluad i the reactor v.sseLlccl.,. , a... ........; specific definitia~n f"r thi:c te:n.., De:-.'pcr Note Thi:c.,, deeipto hul n:d!alteea.ii tu ..., ,,.nac and .poo traugh i.'adiatzed fuel ma' be reaved, but nat including t.he reactcr ;'caacel.RUPTURE(D):

The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.

Dev.eloer~:

Nate Thic tem ic appli.abl..to SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

Dev!elper" Nate Thi3 tenn" may' b'e madified to include the SECURITY CONDITION:

Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.B-3 tUNISOLABLE:

An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED:

A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown.VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.

The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

B-4 HU7 ECL: Notification of Unusual Event Initiating Condition:

Other conditions exist which in the judgment of the emergency director warrant declaration of a (NO)UE.Operating Mode Applicability:

All Emergency Action Levels: (1) Other conditions exist which in the judgment of the emergency director indicate that events are in progr'ess or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist wieh-thai are believed by the emergency director to fall under the emergency classification level description for a NOUE.138 J4,8$YSTEM MALFUNCTION ICS/EALS GENERAL SITE AREA EMREC EEGNYALERT UNUSUAL EVENT SG1 Prolonged loss of SSI Loss ofall offsite SAI Loss of all but one SUI Loss of all offsite all offsite and all onsite and all onsite AC power AC power source to AC power capability to AC power to emergency to emergency buses for emergency buses for 15 emergency buses for 15 buses. 15 minutes or longer. minutes or longer. minutes or longer.Op. Modes: Power Op. Modes: Power Op. Modes: Power Op. Modes: Power Operation, Startup, Hot Operation, Startup, Hot Operation, Startup, Hot Operation, Startup, Hot Standby, Hot Shutdown Standby, Hot Shutdown Standby, Hot Shutdown Standby, Hot Shutdown SA2 UNPLANNED SU2 UNPLANNED loss of Control Room loss of Control Room indications for 15 indications for 15 minutes or longer with a minutes or longer.significant transient in Op. Modes: Power progress.

Operation, Startup, Hot Op. Modes: Power Standby, Hot Shutdown Operation, Startup, Hot_________________Standby, Hot Shutdown SU3 Reactor coolant activity greater than Technical Specification allowable limits.Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU4 RCS leakage for 15 minutes or longer.Op. Modes: Power Operation, Startup, Hot________________Standby, Hot Shutdown SS5 Inability to SA5 Automatic or SU5 Automatic or shutdown the reactor manual trip fails to manual trip fails to causing a challenge to shutdown the reactor, and shutdown the reactor.core cooling or RCS heat subsequent manual Op. Modes: Power removal, actions taken at the Operation Op. Modes: Power reactor control consoles Operation are not successful in shutting down the reactor.Op. Modes: Power__________________

_________________Operation

___________

139 GENERAL SITE AREA EMREC EEGNYALERT UNUSUAL EVENT SU6 Loss of all onsite or offsite communications capabilities.

Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SU7 Failure to isolate containment or loss of containment pressure control.-[P-WR{

Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown SG8 Loss of all AC SS8 Loss of all Vital and Vital DC power DC power for 15 minutes sources for 15 minutes or or longer.longer. Op. Modes: Power Op. Modes: Power Operation, Startup, Hot Operation, Startup, Hot Standby, Hot Shutdown Standby, Hot Shutdown SA9 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Op. Modes: Power Operation, Startup, Hot Standby, Hot Shutdown 140 SG1 ECL: General Emergency Initiating Condition:

Prolonged loss of all offsite and all onsite AC power to emergency buses.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director sheui4dxill declare the General Emergency promptly upon determining that (sie-pee hours-) has been exceeded, or will likely be exceeded.(1) a. Loss of ALL offsite and ALL onsite AC power to specific emerge.ncy AND b. EITHER of the following:

• Restoration of at least one AC emergency bus in less than speeifel hours:) is not likely.* ...... .(S.it:* indic.-tizn cfon'Basis: This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers.

In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should-wilI require declaration of a General Emergency prior to meeting the thresholds for IC FGI. This will allow additional time for implementation of offsite protective actions.Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.The estimate for restoring at least one emergency bus she~*1d-wvill be based on a realistic appraisal of the situation.

Mitigation actions with a low probability of success eheul4-witl not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.i~¶19~5TtntQmWi~

141 that._ cupp~ly pc;wzr tc. thc elcctrical di.c.t:r:ib:u.ti.c.

cyctem that po-ve.-c SAFETY SYSTEMSE.

Th.c.rc.ic yial I, zmzrg.ency but; pr train: of S.AFETY SYSTEMS.....iz ... H+ ...baetkut coping an"aly:°ic prform:'ed in accrdanzz+

wiqth 10 CFR. 50.63 rand PRegulato.ry G'uic, 1. 155, Stationx B!azc!.o:.

Site spceific indication of an' inability3 to adequetey rez' heat from thc core:[PUll1 !nocrt sire cpoeiic v'alues for an incor'e/carc exit thz,.rmozouplz tcmeraturc

°andor irPotr-men~tatin that cannot mneasure do--n to approoiroaely thc middle ofr aeti;'e f"uel ohou.:d u+z the !cv.zzt onr scal rcoding that ic not "ab-ve the top of acti'r fu".cl. If the !l',- et en sLereadin~g ic abov... the top o.f activ'e fuel, +the a........

r.at.. ccc .,vc , ...u. chou.ld not be. incud,..F,.r planto th.+ hav... implemented Westinghouse Orwnere Group tEm.........

pReoonce Gui.:d.eline cntcr the par-t-ro uted nteCr e C....... ling... Re Path....ECL Aoionn~ent Attributeo:

3. l.1.B 142 SG8 ECL: General Emergency Initiating Condition:

Loss of all AC and vital DC power sources for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director sheuildvd ii declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. Loss of ALL oflfsite and ALL onsite AC power to BO11 )All~-': site specific

......uses) for 15 minutes or longer.AND b. Indicated voltage is less than (ies if b ;ztgvauAn ALL for 15 minutes or longer.Basis: This IC addresses a concurrent and prolonged loss of both AC and vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.Fifteen minutes --az selected ais th.e threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.'Dnepe'e-rNetes:

The eitec speeifie em'.ergency buses" are the buses fed by' oft'tte or em..ergncyc3 AC peWer scures~The. 'zite ......i ;b, c v..e tage "'alu'" eh-"uld b.^ b.aced en the hu:s ;vltage neccc=a.-y for adequate e~peratiea af SAFETY SYSTEM. equipm-e~nt.

This ;veltage ;'aluae seheu!d the cd.The v-altue .,i .........

nea the minimum. : ....a.. sel..t. when. barte- c i.- .....ee the. cell.. veltage1..

is apprax:matel 1.7 'J.... t e~t per e]cU. For a" 5° string ba..e.. th.e minimum= voltage is appraximately 1.81 Vclts per ol....143 Th,. -i;t. zp~cific Vita.l DC,-, ...... a thc, DC b. .... tha ........ ; cn i -;,.n... an ...... I fcr SAFETY $YSTEMS.LLL AZ~IgnmCflt Attrltutzz: .j. I.1.Ii 144 SSl ECL: Site Area Emergency Initiating Condition:

Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director i declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(I) Loss of ALL oflfsite and ALL onsite AC power to (site .....ci.... e ...: mergen.cy' 2)AAQ2 1()B +for 15 minutes or longer.Basis: This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.Fifteen minutes ;v'as ais th.. threshold to exclude transient or momentary power losses.Escalation of the emergency classification level weald-beauscs ICs RGI, FGI or SG1.minimu n.. umbr of cpernting+

generators e sr f... r. that source........

pr'vi adequate power to+i.e...to5%cpct g,+ne_ t..r ;+.i......t.

fced I+ e.+ mergeny bus, te[Lan ai seto utseiyta ehgnrtr o h creacceaig The. site spciic+,, emergencyl~

buses are the buse fcd by offsite or emergency AC po +.er sour that suplyl, power+ to the+ electrical ditributio., system. that, l-,wcrs+ SAFETY

.Theretyicll Il emegenc train +IVI of SAFETY SYSTEM UV =UII Th.[A an/o .. as....s. sectio may. .pe.ify u..e -f a n..n aet r.. lated. po, sou.+:-rc++-':e+

prvided that iperotion of,. this source ., 1.++ is contr.lle..in acodac wit abnora or. emergenc.

oprai 145:; + + + + ;++++++;++

+++ + + + + + + 1 +

crcccdur~&.

dc. Iicnt that fla~c a t~rzccduraiIzcd ca~atiIitv tc ~uDch dttzItc AC ~cwcr tc ~e#I~ncd unit via a crz~ tic t~lici that thc pI~r.c~ crc a zn*a ni a cci thi n.. * ,th ac tic ~tmtcgy mcct thc rc~1uircmcnt~

cf 10~R-~O~63~ECL Mcignmcnt Attributc~:

3.1.3.8 146 SS5 ECL: Site Area Emergency Initiating Condition:

Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal.Operating Mode Applicability:

Power Operation Emergency Action Levels: (1) a. An automatic or manual trip did not shutdown the reactor.AND b. All manual actions to shutdown the reactor have been unsuccessful.

AND c. E ITHER of the following conditions exist:* specific indication of an to adeouatelv remove heat from the RC~)Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs.

This is appropriate i~-thabccausc the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Escalation of the emergency classification level ;weuid be- 'auses IC RGI or FGI.po--er le-vel that is: !ess, than or equal to the reactor po--er la-vel which defines the low:.er be'und ot" 147 Applicab~iit.

Far example, if the rcaatcr i.. c.n.idered ta be .shu.tda-,n at 3% an~d Pa-,-er Oeraticno, strt .t....thn.h......a....picb i n Start.p...d.

Dev..... r may.... in'lude site specific EOP crite-ia indic~ativez sfa successful

....ct.r hut.da....

in an EA tat cmo ... the B. is beth (e.g., a reactar pewver leel..." Site specific indication af an inability ta adequately rema-ve heat from the care: tPI4R

.. ite. s..'pecific v'alues fcr an inccre/care therm:cocuple temperature and.'cr rFor vessel whate lavel that drives Wentryinthcore Ow alner procedure (rchrv exit thennacouple" tem.raure great:er th.ana::ae'

-,20an.'ar ater vl ha!o ,spn ds2A ' tcePW apxitely, thcen Pmidl:eof .activer fu~el. Pat ihratrv eee EC L A,'ribuates:

3.1!.3.B 148 SS8 ECL: Site Area Emergency Initiating Condition:

Loss of all vital DC power for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director ehe~ldwilI declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Indicated vote isls hn(ieseii u ctg au LL ~I 12- ) forl15 minutes or longer.Basis: This IC addresses a loss of vital DC power whieh-that compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.Fiftee minues ..a...l.t....

ais the threshold to exclude transient or momentary power losses.Escalation of the emergency classification level -,.'ild be-v4auses ICs RG1, FGI or SG8..:c :c ;pe::s u:.':=o;au aacne a; :: une tha:n'muml:;

.:vs ~ge ne.......

..... fe..r' operation cf SAFETY SYSTEM equipment.

Thi's .. vcltage '.alue shoeuld inearpe-.t n+ .... margin ef at least 15 minutes of operation b.efere. ...h.e enset of inability to operate thee lads. This voltage is usually near th-e minimum voltage selctd he b.attery ......

• i*Th.. typi.cal -'lue for.an.ent.r..b.t... .et 105 .DC .Fcr 60 u cll trin eof batteries, the,. c,.ll .,cl tge .. is appr--imat-l-

.7;.. 5 Volts per cell. a" 50 string batter. set, the...control capabilities for SAFETY SYSTEMS.ECL A s r...........

ribut......

.l... ' B 149 SA1 ECL: Alert Initiating Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director sheu~dwill declare the Alert promptly upon determining that 15I minutes has been exceeded, or will likely be exceeded.(1) a. AC power capability to spezitic cmzrgenqcy bu:;e;)lOTlI I ) I 2 is reduced to a single power source for 15 minue r longer,...

AND b. Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.7*~: Im~pcy~T~wm5 y~s~Th~w~nz Basis: SA\FVIY SYSI 'M: A systcen required lb)r sal'e plant op~ration, cooling down thc plant and/or placing it in the cold shutdown condiiion, including the [I('(S. 'lhcse are tvpficaill systems classified as safety-rclatcd.

This IC describes a significant degradation of offsite and onsite AC power sources (see lFable S I abo'.e) here any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

This IC provides an escalation path from IC SU1.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

150

• A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from the unit main generator.

• A loss of emergency power sources (e.g., onsite diesel generators) with a single train of emergency buses being back-fed from an offsite power source.Fifteen minutes sel'ected as a is the threshold to exclude transient or momentary losses of power.Escalation of the emergency classification level w:ecu!d-be-*iauses IC SS1.minimum. number of

+ generators n..e..a..

for tat our..... to pro.id. required.

po ..... t an A~Ca!, emerg.n.y.bu..

Far xmlifabcu pwer trurac iscomArrsc of two enerator Deveclopers ashuld m,,odily the bulleted ersamples pros'idedl in the basis seetirn, above, as needed Incoming.....

powr .ines) c ompis thr sepa...te.p..e...our.es.

Independence may be determ~ined fro a... reie of.. site UFA, SP or.. , los of electrical pow'er The. EAL

... Bai s ..ti .may speci' u...of a non.. sa.... related pow;er source prcvided At multi unit stations, the EALs may credit compensatory measu,-es that are proceduralized and'.sing generators,..

+- other:_ pe soures ,escibe gs anr al or emergen- cy~ op resting .........

s,. et. Plnt thato .++ hav av procedura ied capability t supply..+

offse AC power to an ECL Assignment Aftrbutes:

3. l.2.B 151 SA2 ECL: Alert Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: SNote: The emergency director sheo4will declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the control room for 15 minutes or longer.Reactor Power RCS Level RCS Pressure In-Core/Core Exit Temperature Wide Ranse_ Levels in at least (Ite ... .........:r... steam generators Steam Generator Main or Auxiliary er Emerge+acy-Feed Water Flow AND b. ANY of the following transient events in progress.* Automatic or manual runback greater than 25% thermal reactor power* Electrical load rejection greater than 25% full electrical load* Reactor trip*

actuation Basis: UNIPIANN I-D: A parameter change or an event thai is not I) the result of an intended evolution or 2) an expeeted plant response to a transient, The cause of the parameter change or event may, be known or This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the control room. During this condition, the margin to a potential fission product barrier challenge is reduced. It t-has-represents a potential substantial degradation in the level of plant safety-of-th As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the control room. This situation would require a loss of all of the control room sources for the given parameter(s).

For example, the reactor power level cannot be determined from any analog, digital and recorder source within the control room.152 An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.

The event wotl-beis reported if it significantly impaired the capability to perform emergency assessments.

In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency pian implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.

In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control hoard, the SPDS or the plant computer, theni the availability of other parameter values may be compromised as well.Fifteen minutes -waz soeacced ae a.ir, the threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level ICs FSl1 or IC RSI1.cr ~Noee: Iii-4he-PWPD carameter ILt eelumn. the ~eite er.ecitie number eheuld reflect the t~inimum num .. ....r .f.am g n...... .. .. .... ..x.....a ...f r .l ..t ......n. ... .... ew........

.... ... ..m.y...epecif whether the ........lu .......d.b ...id .rage ........ .. ran ..r ......th ........g u:Fn th..antr.' for P.CS Level.The. number,

....ie anda laycu.:t ef Centre! R.e.am- ..nd.ca..en.,:':*^-

and the range of pez.eible peio available.....f..r emrgen.y c!oeeifieatio~n a~eczmcnts.

if."- zpecific porer..n:g ef indi!catiene By° fo.u.ing ..n. the avallaiit of the, spcie pa... amet..r...alu..., in.t.a...f.th............th..

vaue,m the. re...i. e an... ac.......ate the. wid. ...i.. ty' of indicatio~ns in nuclear pow;er A loss ef plant annu'.nciators will be for reportability in .accerdance wi.th 1 0 CFR 50722 capatuitit to p~ertor,'m cemerncn.

areseesments.

te..om.pe.salery icr a" less or 153 annunciation can be reodily implemented and may include increased monitoring of main control boardo and more frequent plant roundc by mm licenced operatorc.

Their alerting function notwithotanding, annunciatoro do not provide the pan meter values or specific compOncnt status infcrmation used to operate the plant. or process through AOPs or EOPo. Based on these considerations, a losa of ~~unciation is considered

~o be adequately addressed by repottability criteria, and therefore net included in this IC and EAL.With respect to establishing event severity, the response to a Icos of radiation monitoring data (e.g., process or effluent monitor values) is eonsider~d to be adequately bounded by the requirement of 10 CFR 50.72 (and associated guidanee in NUREG 1022). The reporting of this event will ensure adequate plant staff and NRC a ~ and drive the establishment of appropriate ecmpensatcry measuree and corrective actions. In addition, a loss of radiation monitoring data, by itself, is not a precursor to a more significant event.Personnel at sites that have a Failure Modes and Effeets Analysis (FMEA) included within the design basis of a digital l&C system should consider the FMEA information when developing their site specific EALs.Due to changes in the configurations of SAFETY SYSTEMS, including associated instrumentation and indications, during the cold shutdown, refueling, and defucled modes, no analogous IC is included fcr these modes of operation.

ECL Assianment Attributes:

3.1.2.13 154 SA5 ECL: Alert Initiating Condition:

Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor.Operating Mode Applicability:

Power Operation Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Emergency Action Levels: (1) a. An automatic or manual trip did not shutdown the reactor.AND b. Manual actions taken at the reactor control consoles are not successful in shutting down the reactor.Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful.

This condition represents an actual or potential substantial degradation of the level of plant safety-ef-the An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS.A manual action at the reactor control consoles is any operator action, or set of actions, ie that causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip). This action does not include manually driving in control rods or implementation of boron injection strategies.

If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers).

Actions taken at back-panels or other locations within the control room, or any location outside the control room, are not considered to be "at the reactor control consoles".

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC SS5. Depending upon plant responses and symptoms, escalation is also possible via IC FS 1. Absent the plant conditions needed to meet either IC SS5 or FS 1, an Alert declaration is appropriate for this event.155 It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.the pewer !t-'e] at* wh*ich the reactor ic cneid!ered

eh-tdo'wn.

A PWPR ":.ith a ohu.:tdo,:,:

reactoer Opwrlcl" that+ i,,tc- ++ -.,o/_ +th_ or equal to th raco pp wrl!lwi! eic h lwrbuda..wer.Oper.tion.(M .de .)..ill.nee..t..in.l..

Starup...

d 2 ......in.. O ..r..in .Mode,, term (e +.:.g.. man onr A boardo)156 SA9 ECL: Alert Initiating Condition:

Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1) a. The occurrence of ANY of the following hazardous events:* Seismic event (earthquake)

• Internal or external flooding event* High winds or tornado strike* FIRE* EXPLOSION* Other events with similar hazard characteristics as determined by the Shift Manager AND b. EITHER of the following:

• Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.* The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.Basis: I:IRE: (/ontbustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation ot flame is prefe~rred but is NOTI required if large quantities ol smoke anod heat are- observed.E;XPLOSION:

A rapid, violent and catatstrophic fatilure of a piece of equipmenlt due to combustion, chemical reaction or overpressuri/ation.

A release of steam (from high energy lines or components) or an electrical component tailure (caused by short circuits, grounding, arcing.etc.) should not automatically be considered an explosion.

Such events may require a post-event inspection to determine if' the attributes of an explosion are present.SA:E [IV SYS [EM: A system required (or safe plant operation, cooling down the plant and/or placing it in the cold shutdown condlition, incltuding the FCC'S. These arc typically systems classified as saetyt-related.

VISIBLE DAMAGE: Damage to a component or structure that is readily observable ,avithout measurements, testing, or analy sis. t'he visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affecoted component or structureDamage-t'o a ...........

c :-tinig SAFETYt' SYSTEMS v'zual imac ........ cau ., a.b abou th..a zperabi!ity, af SAFETY SYSTEMSa within: .t. ........ u .... Eapz ,i pa..tial ......a, 157

~cii'a~e e3t er a ..........

cngu--d in t+'~~ ..... nzt includ..e ..... su......t,. nz ....tz zr -- --caizc nr.tc,+:.

splling,h

..... I.,fcrmnt.n

...

Ti Lin--c This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or a structure containing SAFETY SYSTEM components, needed for the current operating mode. This condition significantly reduces the margin to a loss or potential loss of a fission product barrier, and therefore represents an actual or potential substantial degradation of the level of plant safety[hc first threshold tor EAL l.b4-I addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available.

The indications of degraded performance shudwilt be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.fh.c second threshold for EAL I b- addresses damage to a SAFETY SYSTEM component that is not in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.

Operators will make this determination based on....~i~..... :hll available event and damage report information.

This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage.Escalation of the emergency classification level w;eu!d-be-v4auses IC FS 1 or RS I.Nuclar p......r plant SAFETY SYSTEMSc are eomprit;ed

f. t... or mere c~eprate .an redundant.

trano of equipment in .......ne with... ote criteri:.ECL Accie-nment Attribu+teo:

3.1.2.13 158 SUl ECL: Notification of Unusual Event Initiating Condition:

Loss of all offsite AC power capability to emergency buses for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director sheutidwill declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss of ALL offsite AC power capability to (site spezific emergenczy buzeg)Bl~J'l___I

~ ~for 15 minutes or longer.Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources (sce table$2 c) renders the plant more vulnerable to a complete loss of power to AC emergency buses. This condition represents a potential reduction in the level of plant safety- fhe...eplant+.

For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergency buses, whether or not the buses are powered from it.Fiftee minues ..........

d.n ais the threshold to exclude transient or momentary losses of offsite power.Escalation of the emergency classification level we**l-b-..,se

.C.S.l.ic typically.....

1 --* emergenc .......u.. per train a~f SAFETY SYSTEMS.can be implemented "w;ithin 15 mi:n....ute Canidr ...... ac p....................,:

159

... .t .....LJLA~xgnmcntAttntutc~:

J2.I.A 160 SU2 ECL: Notification of Unusual Event Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the control room for 15 minutes or longer.Reactor Power RCS Level RCS Pressure In-Core/Core Exit Temperature Wide Range Levels in at least (1tte vp.-ifx. ... ...r,umb cin steam generators Steam Generator Main or Auxiliary ef EmrecyFe Water Flow Basis:.iNPLANNFl[):

A paramecter change or an event is not I) the resutll of an intended exolution or 2) an expected platnt response to a transient.

The cause of the parameter ehange or event ma'.be knlox\ n or ulnknloVxFn This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the control room. This condition is a precursor to a more significant event and represents a potential degradation in the level of plant As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the control room. This situation would require a loss ofall of the control room sources for the given parameter(s).

For example, the reactor power level cannot be determined from any analog, digital and recorder source within the control room.An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.

The event welbis reported if it significantly impairsed-the capability to perform emergency assessments~.--, particularly thoseln e,,m.rgc ......assessmeents necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, core cooling and RCS heat removal. The loss of the ability to 161 determine one or more of these parameters from within the control room is considered to be more significant than simply a reportable condition.

In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, then the availability of other parameter values may be compromised as well.Fifteen minutes wac seleeted ac ais the threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level ";;ukl-be-v4auses IC SA2.In the PWR param.eter

!izt cclmn t...he ..it. z* pecif number" should refec t..he minimum.number of steam generaterc neeesza&'

fcr plant cooldewn.-

and shutda-;n.

This may. noce Dev*elepers may specify either pressurizer ar reactor .,essel..1 l...l in th.e- ÷1-,WR. paI'rame.

,. ....lum .. ..entry for P.CS Level.fallure challenge the abiit c f an eperater t .... ac atl d..termin...

wihi the.tim.period av'allable for emergency classitication assessments, if a specitic percentage af indications have been lost. Th.e approach used in this EAL facilitates prompt eand accurate emergeney classifloatian assessments by f ..u..ing on the indiatiensfor a .elect. sub... t o fa B"fcusaingi a th e afac'm,~ ef. they sp~eci..iedt paompenster' vales,-insta ofr th !saus of the vnaluensnuedtaoert the EA rcgnze ndo aproedsts t~hecg AOPe arietyPaf iniation in nuhearpse plont CenrolRo niationsatys opcs -:.d andi cos'ereed ma: be anal"og. *er diaddesafet y relcated or net........m....

p..a.... ltenat..idi.dua mete .a.u. :ar.a. c eptr adequp ipate,) ate. (rnd te asentci, a-. (n sated guianeinNUE 122,and c r "nre if it, significantl impair.. 162 thcir zitc EALs.oar~lc~us IC is inclu-ad fcr modes zf cvrat.icn., ECL Aai~nmznt Atthbutc~:

3.1.1 .A 163 SU3 ECL: Notification of Unusual Event Initiating Condition:

Reactor coolant activity greater than Technical Specification allowable limits.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2)(2) RCS coolant sample acetitil value iladicatinlg

[hi clad degradation greater than Technical Specihication allov\,able limibs as indicated bAYof the tbllowing:

~rA~R. 6&A~ItY limit apecitica in I ccnnieal ~pecitieatian~.

Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications.

This condition is a precursor to a more significant event and represents a potential degradation of the level of plant safety-ef Escalation of the emergency classification level weuld -be- auc ...F.orte.ecgito Category R ICs.-* An installed radiati.en monitor on the letdo~e.ve:;tern or air ejecter.The manitar reading "valu:ec should cofe'esp-.nd te. an P.CS nect-vity level apprevimately at....ch.nical ... Specifi atAn; Talla ab le

..lk. .. .. 31'164 ECL ,ccignmcnt At",ributc;:

3.1.1! .A an'd 3.1!. !.B 165 SU4 ECL: Notification of Unusual Event Initiating Condition:

RCS leakage for 15 minutes or longer.Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2 or 3)Note: The emergency director shoutld~x ill declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(I) RCS unidentified or pressure boundary leakage greater than ;'aluz)1! for 15 minutes or longer.(2) RCS identified leakage greater than sp..ifi" value)2$ 15 minutes or longer. [(3) Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer.Basis: This IC addresses RCS leakage whieh-ffayihat could be a precursor to a more significant event.In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of phlan safety of the plat.EAL #1 and EAL #2 are focused on a loss of mass from the RCS due to "unidentified leakage","pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications).

EAL #3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus-apply to leakage into the containment, a secondary-side system (e.g., steam generator tube leakage in a PWR) or a location outside of containment.

The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications.

Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation).

EAL #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.The release of mass from the RCS due to the as-designed/expected operation of a relief valve does not warrant an emergency classification.

Fer' PWPs,-a~n emergency classification weo.1drequired if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.Escalation of the emergency classification level wc, uIdbe-v4auscs ICs of Recognition Category R or F.166 9~op~Ne#eM EAL #1 Far tha ~ita ~pccific Icak rata ;aluc, antar tha highar af 10 gprn ar tha ;'aluc cpccifizd in tha ~itc~ Tcchni~a1 Spccificaticnz for this ~ af1z~o~c.EAL 1t2 Far tha ~ita ~pacific Icak ra~ ;'aluc, antar tha highar af 25 gprn ar tha valuc ~pcaifizd~n tha ~itc'~ Tachnizal Spccificatiznz for thiz ty~c of Icakaga.For zitz~ that ha~'z Tczhnizal Spacificatiana that Ja nat ~pczify a Icakaga typa fcr ztaam gancmtar tuba Icakaga, dc~'ctapcr~

zhauld inaluda an EAL far tuba Icakagz grcatcr than 25 gpm for 15 minutca or longer.ECL Az~ignmcnt Attributa~:

3.1.I.A 167 SU5 ECL: Notification of Unusual Event Initiating Condition:

Automatic or manual trip fails to shutdown the reactor.Operating Mode Applicability:

Power Operation Emergency Action Levels: (1 or 2)Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

(1) a. An automatic trip did not shutdown the reactor.AND b. A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.(2) a. A manual trip did not shutdown the reactor.AND b. EITHER of the following:

• A subsequent manual action taken at the reactor control consoles is successful in shutting down the reactor.* A subsequent automatic trip is successful in shutting down the reactor.Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of plant safety-ef the-plant.

Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor trip) using a different switch). Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.168 A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor trip).This action does not include manually driving in control rods or implementation of boron injection strategies.

Actions taken at back-panels or other locations within the control room, or any location outside the control room, are not considered to be "at the reactor control consoles".

The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA5. Depending upon the plant response, escalation is also possible via IC FA1I. Absent the plant conditions needed to meet either IC SA5 or FA I, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.Should a reactor trip signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.* If the signal causes a plant transient that should have included an automatic reactor trip and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and ,heo1d-u ili be evaluated.

• If the signal does not cause a plant transient and the trip failure is determined through other means (e.g., assessment of test results), then this IC and the EALs are not applicable and no classification is warranted.

ThL!ICisa pplicabl in.an ...d ..in ..ich .th ...tu .l.ra ...r ......l......ul.........h Li..L .~.ssIgnment

,'uirllsuses:

i.I.z.r~.169 SU6 ECL: Notification of Unusual Event Initiating Condition:

Loss of all onsite or offsite communications capabilities.

Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (I or 2 or 3)(1) Loss of ALL of the following onsite communication methods: In planft telephones Public address system (2) Loss of ALL of the following ORO communications methods:[INN (F-mcrgcnc)

Notification Netwo(rk)[(ormmercia!

phones (3) Loss of ALL of the following NRC communications methods: I NS un FederaI I clecoinmunical iob S\ siefm IFfS)[Commercial phonties Basis: This IC addresses a significant loss of on-site or offsite communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.This IC ill be assessed only when extraordinary means are being to make communications possible (e.g., use of non-plant, privately owned equipment; relaying of on-site information via individuals or multiple radio transmission points: individuals being sent to offsite locations,-ete.).

EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.

EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration.

The OROs referred to here are the states of Georgia and South ('arolina; Burke ('ountx. Gecorgia:

Aiken ('ount\. South Carolina:

Barnx.elI and Allendale.

South C arolina: and the Sasannah River Site in South Carolina( ~~cprNt~.EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

170 telephonec, page party cyscteme, radic., tct.). Thic licting should in~clude inctalkxl pla~nt methodc u--ced tao perfrm initial emergency to 0R03 ac deccribe'd in t.he te!ephone linac, commercial telephene lirn.s. radioc:, eatellite te.!ephesnec and in...... uaced cemmuni+. .caticac...

+ t tc ehn4....clo 3+Emergecyc Plan, liting should ...........

t..led plant.equipment and cempenents., and..... -. iteme7+'C' ewe an mainaine ind...dua....The......................i..l.y th dedicated..

LUL' Accien....t Attr+ibut,+:

.t 171 SU7 ECL: Notification of Unusual Event Initiating Condition:

Failure to isolate containment or loss of containment pressure control.fP-uJ Operating Mode Applicability:

Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: (1 or 2)(1) a. Failure of containment to isolate when required by an actuation signal.AND b. ALL required penetrations are not closed within 15 minutes of the actuation signal.(2) a. Containment pressure greater than .. pzzific

............

ImW[AND b. Less than 4 ('1I NI In cnoolcr andc one full train of (" NI I

... .... ee.....:e.... r is operating per design for 15 minutes or longer.Basis: This IC addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results in high containment pressure with a concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of plant safety-ef For EAL #1, the containment isolation signal must be generated as the result on an off-normal/accident condition (e.g., a safety injection or high containment pressure);

a failure resulting from testing or maintenance does not warrant classification.

The determination of containment and penetration status -isolated or not isolated -se*dv ill be made in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The 15-minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.EAL #2 addresses a condition where containment pressure is greater than the setpoint at which containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not have automatically started, if possible.

The inability to start the required equipment indicates that containment heat removalidepressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner.This event weouI4d-xill escalate to a Site Area Emergency in accordance with IC FS 1 if there e is a concurrent loss or potential loss of either the Fuel Clad or RCS fission product barriers.172 FnALth #2~ zp!ii. pe a-u[!r"~ to "z tha. E:tuatxti.

ECL .AsIznment Attr:'5utz:

3. !. 1.A.......c.tp z r c rI o , z(...tary Pov.~r Kza~tcr (EFK) ~ie~ign.173 APPENDIX A -ACRONYMS AND ABBREVIATIONS AC ......................................................................

Alternating Current AOP......................................................................

Abnormal Operating Procedure PRM .......................................................................

~. Aragsz Povwr Ra:nge N1,.eter ATWS............................................................

Anticipated Transient Without Scram IL .....................................................................................

....Bin c lz an.... .....lT................................................................ nta~............B'

................................................................................................

B W R ..........................................................

ioling W aktr Reactcr CDE .........................................................................

Committed Dose Equivalent CFR .........................................................................

Code of Federal Regulations CTMT/CNMT

..........

....................................................................

Containment CSF ..............................................................................

Critical Safety Function CSFST..............................................................

Critical Safety Function Status Tree DBA...............................................................................

Design Basis Accident DC ..........................................................................................

Direct Current EAL.............................................................................

Emergency Action Level ECCS..................................................................

Emergency Core Cooling System ECL.....................................................................

Emergency Classification Level I ..........................

.....................

I ncngc nc.x N or i Ilecal ion Nct ~ ork I ]N'...........................................................................I.

Imcu~cncv Nolilicalion Systemn EOF ......................................................................

Emergency Operations Facility EOP ....................................................................

Emergency Operating Procedure EPA...................................................................

Environmental Protection Agency PGIP..................................................................mr ny Plargnv lPrled'-ent Prax.durin PRP................................................................EEltignnry PIov!crme"t!.

Pr-..c~dz-ER..................................................................................Em rc.................

P~d 4Re........

..........

I ..........................................

Aviaution," AP minise r ationk, FEqdMeA..........................................................Fede dral EmrgnciaiinAgementlageoncS.......................................................................

Fc dcralII I Bo reau onfca ionvs t cinai FEMA .......................................

..................................

FeraEmgnyMngmntiAgeniy II,-..............................................................................................Fcl

..................

R..................................................

Final.......a....t

...............

R Fz GE .....................................................................................

General Emergency A-!

................................................................................

it.....................

Li i TL.................................................................

P!........................................................................

Hg ol,: r......H.................................................................

Fl............................i

...................

RC SIC.................................................................................

r.Initiatin

!Conditio ID(().....................................................................dqmtrsO aIonsid Diame te(r(................................

.................................................................

PEEF........

.. .lnhda Plan Exa~minatian af External ELent, (Generic Lectcr 88 20 ISFSI ......................................................

Independent Spent Fuel Storage Installation Keff ............................................................

Effective Neutron Multiplication Factor.................................

..............................................................

1.t.....................................................................................................

Liden OC ...............................................................................Lss f",,°,tAcd.Nt CR .......................................................................................

Main Ccntral Lcp mR..mme, RM...............................................milli-Roengen.Equival nt a NEIX.............................................................................Mi Nutear Energyia Inaltue NR ...................................................................................

mmisio NQAS ...........................................................

MNrhAmrcniopaeDfnstea Limmnd (NO)U m ........m, ..........

re............mREM...................m(Notificatgen Eq Uuival EentMa I:-AR" ...........................................................

NtcrNaIg::crt"rdRscrc cr l NBEI...........................................................................Nulearatn ai EnergIntthutke NPP ...............................................................................

Nulear PoweroPledAnta NRCM...................................................

NufstDs ClarlRegla Co...........xMmisinua OO..................................................................Off-ite.Rspons.Orgaizatin PA......................................................................................Nlartm uPplytected re.....N..............................................................Noti.....cation.....O......nusual...

Evn PG......................................................................................Gidelne

..........

C.................................................................Owner......Controlled......

Are PA ...............................

....................................

Pccs!f,- ...Proad Aeasc A-2 RAIPSA ....................

Prcbabil!ztic Rbak A~ocmz.'n..t

/ Prabablti.'tc, Safety ,*a~o~.anwn PWR ..........................................................................

Pressurized Water Reactor PSIG......................................................................

Pounds per Square Inch Gauge R...................................................................................................

Roentgen dCC............................................................................

Reatar Care i rlti! Cnzaln R(D1 .......................................................................

Reactor (oolant lrain Tanik RCS ..............................................................................

Reactor Coolant System Remn, remn, REM..............................................................

Roentgen Equivalent Man RPS ...........................................................................

Reactor Protection System RPV ..............................................................................

Reactor Pressure Vessel RVLIS ..................................................

Reactor Vessel Level Instrumentation System V/CU .......................................................

Rearctar W.atear Cleanup SA ...........................................................................

Site Area ntcmrgecncx AR ......................................................................................

Saet ...... AS................................................................................

S .SDAteat.ian Syskate SC. ........................................................................................

South Carolina SCBA .............................................................

Self-Contained Breathing Apparatus SG ........................................................................................

Steam Generator SI ..........................................................................................

Safety Injection leJ \ ..................................................................

Sfr*!:a'.atciun Jet CAir I Sjcctor SN ..........................................................................

Southern Nuclear Compan?SPDS ..................................................................

Safety Parameter Display System TD...................................................................................quialen TOAF.................................................................................Seniprof Active Fuelr T F..............................................................................Top....of...Active....

Fue SC ..............................................................................

Technical Suppert Conter VI)C ................................................................................

Volts lirect Current\I( I. ...................................................................

Vogtle F Iccritt ( icncrat in a Plant V()IP ...................................................

oice O~cr lnternet Protoeol W tl I ..........................................................

A asta I loldlup lank OG .........................................................................

Wcatinghzuec Ow-.ncr5 Group A-3 A-2 APPENDIX B -DEFINITIONS The following definitions are taken from Title 10, Code of Federal Regulations, and related regulatory guidance documents.

General Emergency:

Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.Site Area Emergency:

Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; I) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to, equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.Alert: Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION.Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.Notification of Unusual Event Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.

No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.The following are key terms necessary for overall understanding the NEI 99-01 emergency classification scheme.Emergency Action Level (EAL): A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level.Emergency Classification Level (ECL): One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (I) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in -dtcsccnding order of severity, are: General Emergency (GE)Site Area Emergency (SAE)Alert Notification of Unusual Event (NOUE)Fission Product Barrier Threshold:

A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier.B-I Initiating Condition (IC): An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.

Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document.

The definitions of these terms are provided below.CONFINEMENT BOUNDARY:

I1 he barricrls) between areas containing radioaeti\xc substances and th.. enviorii'n.(l"zr a ......i d*:' i " itin fre thit5 tc..rr.) ,a .... r Na.. : ba.ri:r.,)

bu c .........

fu.c, an.d tz ..... n....t zn t... fue"l i; pracc~cd fcr dry, CONTAINMENT CLOSURE: l'cr Operating Procedure 14210-I 2. C ontainment Ituilding P'cuclt iaons Veritication

-Retie in ,~ f a itx~ .: ....... : d. .in.i:, an for.u: thi tc n .) .n .....Noe= The prczedur=ly defined ar taken ta. zae-ra zant.ainmcnt (7primary, r EXPLOSION:

A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.

A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.

Such events may require a post-event inspection to determine if the attributes of an explosion are present.FAULTED: The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to become completely depressurized.

Natc.e Thic t.... : .....~bl ,:t., P,. an... .FIRE: Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute FIRES. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station.HOSTILE ACTION: An act toward a nuclear poWvCr plant (NPP) or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.

HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the NPP. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area (0(A }).HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

B-2 N cI 99 01 29:z !2 IMMINENT:

The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions.INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.OWNER CONTROLLED AREA (OCA): The site property os'ncd hx or under the control of V[IG' .eort... * ... a cte .cpccfic def.n.tien fr. .: thc .nn.) Df-cvzlopr ,Nate -,Thic perimeter to... th.. plant Patcte Ar...a perim.ter ( .... g ... a* cite th a large OC A w:here some deap..... , .... d .on. i ... : u ...i:n g. the b.undary, defned by the Rec.tricted a~r Secured O-w.ner PROJECTILE:

An object directed toward an NPP that could cause concemn for its continued operability, reliability, or personnel safety.PROTECTED AREA (PA): [he area that eneotpa sses, all controlled thinhh the sceurity protected area fcner l nc cr. cite.....

cpeii .................................

p rN .....h ...t..

t...n t. me.... the ......a under .: .............

~menitacring an'd centrel, and a.rmed REFUELING PATHWAY: l his includes the reactor refuel cavil',lhe fuel lran~slkr canal, and the Sl'pnt tue pool. canals and pools through x Ihich irradiated tuc I ma\ he moved, hot n/ot ineluad i the reactor v.sseLlccl.,. , a... ........; specific definitia~n f"r thi:c te:n.., De:-.'pcr Note Thi:c.,, deeipto hul n:d!alteea.ii tu ..., ,,.nac and .poo traugh i.'adiatzed fuel ma' be reaved, but nat including t.he reactcr ;'caacel.RUPTURE(D):

The condition of a steam generator in which primary-to-secondary leakage is of sufficient magnitude to require a safety injection.

Dev.eloer~:

Nate Thic tem ic appli.abl..to SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related.

Dev!elper" Nate Thi3 tenn" may' b'e madified to include the SECURITY CONDITION:

Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A SECURITY CONDITION does not involve a HOSTILE ACTION.B-3 tUNISOLABLE:

An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED:

A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown.VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.

The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.

B-4