ML16071A193

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

ML16071A193
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: ML16071A193 (39)

Text

{{#Wiki_filter:SG1 ECL: General Emergency Initiating Condition: Prolonged loss of all offsite and all onsite AC power to eamergencey\>ssntiuI buses.Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director shot,4d-x ill declare the General Emergency promptly upon determining that sp..ee"I4e hours-) has been exceeded, or will likely be exceeded.(1) a. Loss of ALL offsite and ALL onsite AC power to 160 X/AC en¢tla Buses AND b. EITHER of the following:

• Restoration of at least one AC effirgencycsscntial bus in less than .i-tpee4iqe4 hours-) is not likely.* Recactor xessI x ,,aicr k".cl cannot bc restored aud m~aintineiud athuxc Mdinin um Basis: This IC addresses a prolonged loss of all power sources to AC emeegeineycsscntiaI 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 sh \il-i Il require declaration of a General Emergency prior to meeting the thresholds for IC FGI1. 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 etmefgent) essential 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 emefgeneycs~scnmial bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should 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.131 Aithheugh thie ,C an....... a...... iew.d....r.d.nd....t.. th.............d.....arr..r....,..t ic included ........i:e.fr a.. more timely of the ern.ergeney ,laci.cat:n. , l....l.T....... t-t- ifie encrgte. y...... '...... h......... f.d......... t... r.. m.. rg. n.y......... .......ic typically, .... ..m.rg.n.y b.,u.cL., per train ef SAFETY SYSTEMS.The "citec pecifie houre" te rcctore AC pew;er to a em emrgency buc cheuld be ba~cd cn the Gui'-de 1. 155, St~rioh BIhwckoa:t. Site 3cpcific irndicaticn of an inabfilit te adequately remove: heat from the co-re:........ Rece .ec.....J ..r...;'"' I~dl ,nof b r.to-ed andl maintai.. d ab ......... Miiu S.. am re :a-er thel atep.r le,,v.l tha..ri... ntry.nt..a..r....ingr ......i.n.....ur...r..thrwic reqir .iplamntatiuon ..f _prp gc.i..... Altrnt..l.. a... citeg may uc ........r I A tAiL Acci~nment /'dtributec: I.t.lJi 132 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 sheo4d-ax ill declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. Loss of ALL offsite and ALL onsite AC power to 16 VAE ti 1 1, L.F A specific emergency buzez.) for 15 minutes or longer.AND b. Indicated voltage is less than 05/210 VI ste specific bus -veltage ;alue) on ALL 2/ I 2R2S0l6~J 212-S0O 7(sitc specific Vital DC basses) for 15 minutes or longer.\kU 4 ~AVAt~ 4~b~a~ioz~ 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.Ffenminutes was ...l...t.d as a the threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.133 Th~ 'zitz cifi~ Vital DC bu~z" arc thc DC bu~c~ that pr~;idc rncnitcring a~d zcntrcl capabiliti.z~ for SAFETY SYSTEMS.This IC and EAL wara addad tc Rcvi~izn 6 ta addrca~ zpcrating c~pzrizncc fram tha March, 2011 aczidcnt at Fukuahima Dpi jahi.ECL Aaaignmcnt Attributca: 3.1.1.B 134 SS1 ECL: Site Area Emergency Initiating Condition: Loss of all offsite and all onsite AC power to e ergenecyssenial buses for 15 minutes or longer.Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: SNote: The emergency director §heoltd-\x ill declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss of ALL offsite and ALL onsite AC power to 1.VC i I ZE.i/2F o sit specific ..m.rg.n.y bu....) 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 wes-se eeted-as is a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level wo4-v.,:be -*iauscs ICs RGI, FG1 or SG1.Del~epe~s-rNe.'eM Far a pev.'r source that multiple generators, the EAL and'er Basis scti:en she.'uld reflect the i.e.,.t. 50% capacity generators sized ta feed I AC emergency bus), the EA an, Basis: The "site specific emergency busesz" are the buses fed by effs ite or emergency AC pow:.er sources that supply pewe'r to the electrical distrib.tion .......m that, p....r. SAFEctTY SYSTEMS. is typically I emergenc.. bu per.. trai ef SAFETY SYSTEMS The.EAL.and..r..asi.....t.. nmay u............. saet rel.ted pe ........ prvid....th-at operation of this sc~urce is contrelled.- in .. ..._.aee .... with abnor.mal ar emergency ..p.. rating:-proc.edures.. er beyand design basis acci:dent resense guidelines (e -g., FLE ......... guideies)At multi unit statons the may credit compensatory measur"es that are proceduralied a.nd pr..d.. s etc.... .Plnt ...at have .. a proceduralized te .. suppl ^offsit AC power te an 1354t 60 rn i

• .oA f affe~ kd unit ;Ia a zrZ ~z~zcmr~anxcn unit n~ay ~r~Iit tflIa pzw~r ~curzz in tfl~ L~L:m;'idzd that th.~ vipnncj Crczz ti~ ~trat~zv mcct~ th~ rc~iuzrzzncnt~

CI IU Li-K YJ.t3.ECL ,Azg nmzrnt ,Attributec: 3.!.2.136 SS5 ECL: Site Area Emergency Initiating Condition: Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal.Ope rating Mode Applicability: Power Operation Emergency Action Levels: (!) a. An automatic or manual scram did not shutdown the reactor.AND b. All manual actions to shutdown the reactor have been unsuccessful. AND c. EITHER of the following conditions exist: ai~xjuta!y hemt frcm ccre) ~w Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram 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 tlht~s-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 in that 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 weteld-be-v4*usc-, tC RG1 or FGI.137 De~epe*-Nete~t L I t I fliC Ri iC applieatle in any MaCic in wfliefl tne actual reactar p..,~.. L. ,A .,.,uia exeeco tne.. o! + q I ,I....... I...... Ie'c at.. , h+.c..fl tf. rcac.tor iC caCijeredC huJtd.?;vn.AI". A with .a,*. ,.hutda;.wn rcactar.p....r lee th. t i :cc tha r ....qual ta the. r-eactar paw;;er level whic:h dAfec thc_ .......... d.PwrOprata (Ma"d- I) wil" n... to include. .ta.tu. (.ad 2) in the Opc..t.. :-_ de Aplc;-abiiy.Fratcn, mp..,.if.....r...t.r............- t, ..... zhtn ..t. 3% a.... P.....Operatian-sc'tsz ..t t..n h.e ,C pisals applicable in Sta."tup an EA.L etatement, the.. Bacic: or bath ,-.g.,t" a reactor power level).Site Cpacific indicatian afar. inability Ia adequately remav'e heat fra~m the care:[P;U,. Inpertn ;CJite+ ....fi ., far an nae/uarey reoie thenaauple temperature anoi'requi,,rJeC itperHetatia Caprampt Tectratiurn acmtia. lTeisnatdelysasiete mnayi~;'t ucm'e iae/art ,exit thA niacatpe great.ertha Ia:'e ,20F anda ratr ee.wae evl FCL .Meianment AttributeC: 3.l.3.B 138 SS8 ECL: Site Area Emergency Initiating Condition: Loss of all V4qal-x ital DC power for 15 minutes or longer.Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: SNote: The emergency director sotidd-x~ ill declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Indicated voltage is less than VI sp~cific bus '.ocltage value) on ALL 125i7 VD us1222$I6 fR22 p....i Vi...........l.: DC,. b~-uz.... for 15 minutes or longer.Basis: This IC addresses a loss of vital DC power which 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.Fifteen minutes wis eiiN a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level we.*.uldbe ICs RGI, FGI or SG8. Tha %!te s~pcifk vezltagz value" h be "de zn tha minimum b'-c ;'eltage ncc.......... ............ e-...:' fS.AFETY SySTEM equipmcnt. -Thic...o rtagz ;value chould ince.'boratc a ma.rgin ef at leact 15 m~nutec of z.pemticn hafore- the .....t cf inailty:;. t........thc.e load.;..! Thi " vc ........ ucuall ner th ..minimum ~'la rc. J .. h.n-,,tk c..Izi ezntrc,! capabilitices fcr SA*FETY SYSTEMS.ECIL Acc!gnmrcnt ,Attrib!ut.ce: 3.!.3.B 139 SAl ECL: Alert Initiating Condition: Loss of all but one AC power source to emsergencevessential buses for 15 minutes or longer.Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director declare the Alert promptly upon determining that 15[minutes has been exceeded, or will likely be exceeded.(I) a. AC power capability to 1 A i alD /EIF: .O~~e spedific emergency bu~ae) is reduced to a single power source for 15 minutes or longer.AND b. Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.[73k: Vt~ 4 6Oy~4C P~ses tntqp~ti~I L741VI34I~VACBu~r~tio~i Basis: SA -12k IV y~r SY ,l F: A s?,'stertn required for salte planlt ope~ration., cooling down the plalnt anid/or placing ii fin the cold shutdownl condition. including the LC(iS. These arc typically s~stemns elassitied as siifet y-related. This IC describes a significant degradation of offsite and onsite AC power sources (see '1 able S 1 ) such that 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 SU 1.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergececycssenI ial bus. Some examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all but one eimiegeineyessential power source (e.g., an onsite diesel generator).

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

140

• A loss ofe frefiey.>scntial power sources (e.g., onsite diesel generators) with a single train ofeffiegenecysscnfial buses being back-fed from an offsite power source.Fifteen minutes we-eeee-is a threshold to exclude transient or momentary losses of power.Escalation of the emergency classification level -we~uld-bc viuscs IC SSl1.minimu numb.. r of_ eperteing gen.... r,. n.ce....ary for .th.........eto.pr..ide required power to an AC..m.rgency hu,.. Fo ........ if fa backup p....r.......

ic ......ce ;of ftw- generator.o (it:e..,," no/ 5 .... caact ........c sized. toA fee ,AC-b'us), the* [,AL .ad Basic must'c sprcify that.. bothFc gnrtoree fndr that sorc ar5e oerating.pwe rc! (e.Them rc;ite sp fe st pcificeere cybs" UFSrPe the buse :,sife by of~chitd or: emfrency Cr powz orer.... r hud oiy h ulee sape ro..id..d in. rhebasid se:e cti-n, abo e;*a nded The E.Lc and'e Bosic :chtud eflc that pcoch indepen.n....t....r....t ontiut tha perati..n

- f fthic source ic recognized in A.OPc rand EOPs, or b}eynd decign basis ..c. dentmulti unit" tations.'

the , Eac 3 credit co natoy mredit ur th..t...........d:ralized A n provldeJ that the piannel crocc tic ctrategy meetc the requirements ci U ~K ~ECL Assi~mnent Attribotcc:

3.1.2.8 141 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: Note: The emergency director she*4d-w ill 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 1 5 minutes or longer.Reactor Power RPV Water Level RPV Pressure Primary Containment Pressure-Suppression Pool Level Suppression Pool Temperature 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 scram* ECCS-{S-) actuation* Thermal power oscillations greater than 10%Basis: l NPI ANNI';!): A parameter change or an evenot that is not 1 ) the result of an intended es olution or 2) an expected plant response to a transient. [he cause of ihe parameter change or evenA may he known~ or ulnkn,,)\,,n. 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 thus represents a potential substantial degradation in the level of plantI safety-of-te 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.142 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 would be 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 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, RiPV level 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 RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.Fifteen minutes. .... ......et...d a o ais the threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level .....itA eus ... s Fs loICR1 In the. P aamecte~r list c~olumn, the.. ..it. scii ... mb..r...........r........t....mini..um num:ber of steam gen ~rat r nc ary-' for plant cooldown; and "shutdown. This crite~rion may also speei!.*' whether lee ... lue .,. o ... be.wid rang,.na.w rng..........

r. b.eth, dpnding up...n th monitorin'g require~ments in emergency epe~rating procedures, De...l.pers may. spwcify ..ither. pres.uri..r.r.....~....

e.... I. in t.e PWRo paramete~r co'-umn The type-, location layout of Control Room indic~atio~ns, and the r... of possible faiur m.d... an chall. nge the.. abiit of:" o" operator to ac~curactly de~termine, within the time period available for emergenc~y classification assessme~nts, i.f a specific perc.entage of indicatioens classification: asessme~nts by focusing on the in:dic~ations for a sel!ected subset of parameters. values, the EAL recoegnizes and accommoda.ctes the wi;'de -variety of i.ndications in nuclea=r power plat Cnrol R... oms... Indiation: soeurcecs ma~c:.' be analog or digital. saft r...e+ate or nopfmi orater -ate,idvdal.mete value or com.pute~r groupdspa.ec A l.o!ss of pla~nt ann:unciators will.. ..* ,.) .. ....e.aluatd for reportabiliry' in accrdan:ce CR, to cerfcrm emero'enc\' assessments. Comp-ensatorv m'.easures fcr a less ef 143 a.nn.unciatienz radily, implemented anzd may in.iudc inereezed menitering ef main centrel.............. r .frqun ..... rend b...... nen ........... eratere. Their alerting fu:netien infocrmatie c .:ed te eprate the plant. er ureeec threzugh AOPe er [OPe. Bae~zd en t.he.c requirercnent of 10 CFR. 50.72 (and .....te guidan.e. e .n.. EG10. ; lllThe repo.. in. ".fthiel.appropriate compeneatery meacures and cerrectire ati:one. ......diti..., aae ..f............ thei:r cite epecifle EAIs.-h refueling, and defueled medee. ne ar.alagoue IC ic included tbr thece mede: efoperatien. EUL Aecienment Attrlbutee: 3.1!2.B 144 SA5 ECL: Alert Initiating Condition: Automatic or manual scram 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 Emergency Action Levels: 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.(I) a. An automatic or manual scram 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 scram 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 plaint safety-ef the pla~n. 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, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). 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". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.The plant response to the failure of an automatic or manual reactor scram 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 RPV water level 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 145 possible via IC FSI. Absent the plant conditions needed to meet either IC SS5 or FSI, an Alert declaration is appropriate for this event.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.Thic IC ic 3pplica.ble in any, in w:hich the a.ctual". rea.cter pew:,er level ceuld exceed Opentien ....... at / ,5, then teiC ic appl:ca.ble in Startup. Made ECL Accignmentn Attributec: 2.1!.2.8 146 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: H"R!: (iombustion characterized hy heat and light. Sources ot smoke such ats slipping drive helts or overheated electrical equ~ipment do not constitute FIRES. Obhservation of flame is preferred hut is NO I required if" large quantities of smoke and heat are observed.FXPI()SION:

A rapid, violent and catastrophic htailure of a piece of equipment due to combustion, chemical reaction or overpressurization A release ot steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing.etc.) should not auttomaticallv he. considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.SAIFE ]Y SYS ['EM: A system required for plant operation. cooling down the planmt and/or placing it in the cold shutdown condition, including the I-CC'S, [hese are typically systems classified as safety-related. VISIBLE DAMAGE: Damage to a component or structure that is readily observable wxithout measurements, testing, or analysis. fhe visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. 147 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[Al. l.a identities haiardous events that could result in damage to plant sy stemas. A seismic c,,cft is inldicalted bS entrV into IC 11,12. [looding is indicated by a significant increase in \xatcr lev els (external or internal). Ilich arc indicated by sustained v, inds at the site meteorological toxsor exceeding 35 mph.The tirst thresho~ld fbr EAL l.b4- 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 should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.[he second threshold tbr EAL l.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 the totality of 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 weald-be-i ,,:e IC...o.RI ECL A~ignment Attributes: 3.l.2.B 148 SU1 ECL: Notification of Unusual Event Initiating Condition: Loss of all offsite AC power capability to ewiergenceye.ssenial buses for 15 minutes or longer.Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director ill declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(i) Loss of ALL offsite AC power capability to Esv 1 s I/2I! 112 i2 f l..it .. sp ....r..nc. btss r 15 minutes or longer.i 1t~ jD IR 2 C7~1: V~ ~ 416Q vAE ~p% Wb~ou~Basis: This IC addresses a prolonged loss of offsite power (see 'labic S2 ahux c. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emer'genec;'ssentia! buses. This condition represents a potential reduction in the level of phcrn safety-e4f-41e plan4.For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergeneycsscntial buses, whether or not the buses are powered from it.Fifteen minutes wa slete a...... ak ihc threshold to exclude transient or momentary losses of offsite power.Escalation of the emergency classification level w;e',dd be-v4auscs IC SAl.The ..i.. sp.cifi. emer....nc.. bues .... t*. he. bue fe.. byA or ...urc..is typically I ..m.rg.n.y per train of SAFETY SYSTEMS At mult unit stations, the EALs may credit compensatory m.-:easures that are procedu-ralized and procedures. etc. Plants that have a proceduralized capabilib" to supply cffsite AC power to an affected unit via a cross tie to a companion unit may credit this pewer source in the EAL prov-ided that the cross tie ......~ met the..... .gire.......f..0..... 5.63....149 ECL Assi.nmcnt Attrib"ut: 3.!.I.A 150 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 shouild-xx ill 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 RPV Water Level RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature Basis:[ NPIANLI)D: A parameter change or an exent that is not 1 thea result of an intended evoluittion or 2) an expected plant response to a transtent, Ihe cause of the parameter change or event ma.be knlowni or 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 safety-ef the plaft.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.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 ,etid4-be-is reported if it significantly i mdi pairs the capability to perform emergency assessments.=- t-nparticularly those necessaty to ;amcrgency acazzmz~ta necezsary ta 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, RPV level 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 151 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 RPV water 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.Ffenminutes w;,-selee ..... si a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level ":e*-Id bc .'4aiscs IC SA2.In the PRWP parameter lis ....u.... th .... te* specific number" s!hculd. retlect the minimum.menitzring requirements in emergency operating procedures. Th....um...r, *ypr ............. a ....u .....C........................ a, th.... range ecfpible fe uemodes, ca ..hall..ng. th ..ility of.- an pc-retorr.---' t...c.u.ately dctcmin: .ithin th. e time: period available for emergency classification assessments, i:f a specific percentage Cf indications have been lost. The app--aech used in this EAL facilitates prompt and acc"urate emergency elassificatian "oss....sments ..y fousn ..n th'e indications for" ao od :ubest of parameLrs--. By. focusing on the- av'ailability, of the ..p.cified-parmr.eter .,alu...es ... intadfth so~u......rce .f.h..plant Con... ,tro .... Indiaticn ....... an ........ may be ana'-g er digital. safety, relte ...oer A less of plant annu-nciators -w.ill b ev'alu.ated fcr repc:rtability' in accordance with 10 CFR 50.72 (oan the .......c.iat guida"nce in NUREG 1022). and repCrtced if it si'nfi:nt' the capability/te perfolem ....ergen.cy o.ass'ssments. CompensateD' measures fcr a Ies5 cf b..rds ;an more. fr... quent. pl" n ....... by:: .... licend 'erator. Thi nC.........d ng .............. dc not..... prvd the... prameter............p.ci.. cmpCn._enttau crniteriaand athaefr nct included,.o ... i... thiusr IC an meresgL. cn ee 152 thzir site zpafc!q~ EAL:.Due tc. chmagc: in thza ccnftguamticr." o~f SAFETY SYSTEMS, inaluding az~cciatad

i...........

a ... and indi;..izr^_. during thc ec!d rcfuezling. and-q dzfu.kd m^dz, na................ Iv U--k., L, / i "gU. / ",,;,;.r[ : &, i. ;.,", 153 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: (I or 2)INote: I Isc the I T nit I or I lnit 2 lPrctrcatrncnt vs. ntRihr) (iraphs to deterinen if thc Pretreatment Radiation Monitor cxcceds the TV o[ 24t).000~ pCiiscc.(2) Sample analysis indicates that reactor coolants sFcitic activity ,ahte-is E! I IIER:* r*tkgink ~ q~ ltif doe lt3 for* raethn.1 e/adoeettia ntnireacrha .al ,al l.mi .ncfe I 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 efthe-plant. Escalation of the emergency classification level weu!d-be-v4aoses ICs FAI or the Recognition Category R ICs.er EA han Ie! En~ter the radiation mcnittrc) thatamaynge--*ith tre adil-ytc idntifryar -wh 'en RC aciiylvl xedTechnica! Specification allowable limit. Ti A a.b eeee!f newre capnbiexitieicng t rzte, qu~aired) E fmpeeo detiet.ing thmEAthhed n paiitic inclu:dne: e!zd 154 F~r EALit2 D~vJ~p~rz may r~wcrd thc EAL tc incluic th~ rcactcr ccalant activity parwnctcr(~) 5pcciflcd in Tcchnical Spcciflcatiznz ~d thc ~cciateJ allzwabk Iimit(~) (c.g., valucz fcr d~zz zguivalent I 131 and grc~z activity, zinc dcpcr.dcnt cr flansicrA va1uz~, ctc.). H thL appr3ach i~ ~clcctcd, all RCS activity allo;vcblc Iimitz &h3uld bc includcd.EC!. .X~i~nmcnt Attributcz: 3.l.1.A and 3.1.I.B 155 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)SNote: The emergency director sheokld-w ill declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) RCS unidentified or pressure boundary leakage greater than gpn 7!3 VJ ~44S ~ ua for ! 5 minutes or longer.(2) RCS identified leakage greater than ,'ite z~pcitic v'alue ,5gp or 15 minutes or longer. jy :r4 (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 which may 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 plant safety-ef the-phsii. 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 UNISOLAIBLE 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. A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.Escalation of the emergency classification level wouldbe v4auses ICs of Recognition Category R or F.156 in th: ;it&': T~chni[ca1 Spzf. tbr this .t~p zf [z-ka;=z.E AL f-, F..or. th : ...... :pcifi '"a' ..... -'a .enter thc highc f2 p rthtu pof~in sitc'h Thzhnica! Spczifiz:.aticn fcr this of for IS zr !Ing~r.157 SU5 ECL: Notification of Unusual Event Initiating Condition: Automatic or manual scram 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 scram 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 scram 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 scram 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 scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram 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-4f4he-plea'..

Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). 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 scram 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 scram using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram 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.158 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 scram). 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". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action,.The plant response to the failure of an automatic or manual reactor scram 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, and other concurrent plant conditions,-ete. 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 FA I. Absent the plant conditions needed to meet either IC SA5 or FAI, 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 scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance be applied.* If the signal causes a plant transient that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and e*dxx ill be evaluated.

• If the signal does not cause a plant transient and the scram 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.-Th:... IC" is applca.ble in n:. ...d..i...hi.h...h........

reactar paewer level c..u'd exceed the..paw"er !-v;cl a.t ,which thc reneta~r ic caencidered chu-tdon-. A PWR --ith reactor anEL-atmnth.ai ........... bt (.g..., .. rece poe rcFcc ;h lvel)."Fr~!t

...L .~-.ccignmern

.'~.nrit~i.aec: .c. 1.1 159 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: (1 or 2 or 3)(1) Loss of ALL of the following onsite communication methods: (2) Loss of ALL of the following ORO communications methods: FNN (Frnergency+ Notification Netx, ork)( onnuercial phones (3) Loss of ALL of the following NRC communications methods: L NS on Federal Telephone S;} sten, FI) (onmmcercal pho nes (zi.te spcci.ie l:st zfcc....... zn methc.... d.,,.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 should be assessed only when extraordinary means are being utilized 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,--~et). 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 thle state of (ieor~ia. Appling County.J lif lDa\ is Connt',., l'aunall Counlty and T oombs Count" '4+ r"... .. .. D vecz .os...x...+. EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. 160 EAL #! Thc~ "shtc spccifc !ist cfccrnmunic-"ticns mcthcds" shsu-ld includc all..................... 1...... d.. .. .......... nt ..... ......... ......s :cg., cs.nm crcial cr sit: tc-p lc s.v ............... c ....... s,......i..l.. tin......lm ine.' c ld..i....... pl..EAL fl2 .The "sit specific list af methads" shauld! incluadc all oom.'municaticn' "',.-Em........ ,1-hc.. listin';*;g sheuld include, ir~stall.d pl a.nt cguiprncnt and ccmpncnt, s, and net^iterns awnad an maintalned by, Era-mpl: metheds rin.g dw-:,'d:dicoated cemmunicatiens techneelegy. typically" within= 15 minute;.Emergency The li.tin. shul ....l...............d pl..t cquiprne-nt an.... ce .......... an..n.t items e.w-ned a.nd iaintained b) These m'ethods tpically' the dedicated... ........................... .l [Il.C 161 APPENDIX A -ACRONYMS AND ABBREVIATIONS AC .................................................................................... Alternating Current AOP ..................................................................... Abnormal Operating Procedure ATWS.....................................................................Aniiae Tvransen PWitou S~n c AW ................................................................................ ou BScra&W............................................................................. Babcockc and Wilcc'................................................................................................. B liT Bcc .... ticn_,_ !nitiatian Tc.m.pcraturc lBl l)(G.............................. .................. Building, BWR .............................................................................. Boiling Water Reactor (1B .............................................................. C onirol 1B uilding CC.......... .......................................................................... uhie Centimeter CDE......................................................................... Committed Dose Equivalent CFR......................................................................... Code of Federal Regulations ................................................................................................ C SFS Criti"a Safety Functia .aw Ic................................................................................... Cint .... cidr~(P .................................................................................. (....u......P.r..Minu...e (SFT.........................................................................(Couc!SfeyF ncts er Second re..................................................................................... Direct.....Current.. BA 1................................................................................. Doseg Iqialent Iodiet (1 ........................................................................................ D nl r Minuel EAL................................................................................Emergecy Actioalnt Levele ECCS RM..............................................................Emergncyl CodeRae Cooling Syste ECL .......................................................................... Emrec ClassiencyAction Level ECS....................................................................lEmergency NotifCotiing Sytstem EL 2...................................................................... Emergency Noassification Levtel EPA.............................................................Environmental.........Protection......Ag .Enc............ PG...............................................................Emerency, Proedue .Gidein PIP ................................................................. Em Irg c rgcy Pl Notification Networkr PNR ...................................................................... Emcrutinar Povifcazr SRscacz EPA ............................................... Envctiro m na Pcw rotection Agncyiut ERG..................................................................EEegnyPa mpcrgenc'y: Prc~cpodu G AXviation Administration A-1 t:B ...................... 1.............................lederal lBureaui of Investigation FEMA ....................................................... Federal Emergency Management Agency FSAR....................................................................... Final Safety Analysis Report Il ..................................................................... Federa~l l'elecoffmunllieationls Systemt GE .................................................................................... General Emergency iM ....................................................................................................... (iram HCTL................................................................. Heat Capacity Temperature Limit INI P.....................................................................................I latch N uc ear Plant HO ................................................................ Hlcadquartcrs Operations Officer (NRC)HPCI ................................................................... High Pressure Coolant Injection................................................................................................ H S! Hum..n Sy':tam Interface IC ......................................................................................... Initiating Condition................................................................................................. ir PEF.......Indi,,idu.. Plant Eam..t.. af,: Ext ..rn..l E-ventsq (G.enearic Lette,;r 20)ISFSI...................................................... Independent Spent Fuel Storage Installation ................................. ............................................................... K................................................................................... n ...................... Fa t CO...........................................................................Liitn Canditian if Opemtien LOCA .......................................................................... Loss of Coolant Accident................................................................................................. CR.................................................................................. Mi a'r......N MSL ..................................................................................... Main Steam Line t i.......................................................................................n m cro-Cturic mR, moRero, toremo, mREM ........................................... milli-Roentgen Equivalent Man W ............................................................................................... g.-Nat NI!"................................................................................................ Northeasi NEI............................................................................... Nuclear Energy Institute NPP.................................................................................. Nuclear Power Plant NRC ................................................................... Nuclear Regulatory Commission ................................................................................................ N NORAD............................................... North American Aerospace Defense Command (NO)UE ................................................................ (Notification Of) Unusual Event............. .................... ............................................................... N..................................... Nuclear Nlanagcmcnt, and Rcacurcce Ceuncil OBE.......... ................................................................ Operating Basis Earthquake OCA.............................................................................. Owner Controlled Area ODCM ........................................ Offsite Dose Calculation (ssesm........-Manual ORO..................................................................... Off-site Response Organization PA .......................................................................................... Protected Area A-2 ACS................................................................ Prierity ...Actu.....t ...... .. an Cnr,... cal ..... ....PAG.......................................................................... Protective Action Guideline I1X ................................................................................ PIrivate Branch Fixchange P S.................................................................................Pn Srstc ICS .......................................... Przceca and Ce~ntral Syctam PRA/PSA .......................... Probabilistic Risk Assessment!/ Probabilistic Safety Assessment ................................................................................................ P WR............................................................................. Prcc:'-rized Water Reactor................................................................................................ P S ................................................................................. Prtection Syatem PSIG...................................................................... Pounds per Square Inch Gauge R................................................................................................... Roentgen RCIC .................................................................... Reactor Core isolation Cooling RCS .............................................................................. Reactor Coolant System Remn, remn, REM.............................................................. Roentgen Equivalent Man RI1R .............................................................................. Residual I leat Rcmo\ al RPS ........................................................................... Reactor Protection System RPV .............................................................................. Reactor Pressure Vessel VLIS....................................................P.Reeactr V'esce! LeveI !netrumentatian Systemn RWCU............................................................................ Reactor Water Cleanup Rx ...................................................................................................... Reactor SA(U..........................................................................

......eer Accident (inidel inc SAR...............................................................................

Safety Analysis Report A ................................................................................ ...A S ............................................................. a........ itr an S!actern'N.............................................................................Secondry..o.....Sn SCA......................................................................Stl-Cnane ratling Apparautu................................................................................. nd... r......nt....nmen..t GSC.................................A................................................Sl-otie Breathn Apparatusr ................................................................................................ S S P.......................................................................................Sta Senaratar SF ....................................................................................... SlPoo SC......................................................................................ompan SPS ................................................................................ SaeyPretrDslySyuheste .................................................................................................. r......r SW\ .................................................................................................. South west A-3 TEDE................................................................... Total Effective Dose Equivalent TOAF.................................................................................. Top of Active Fuel IV ........................................................................................ lhreshold Value VAC' .......................................................................... Volts Alternating Ctua'nln l)irecct CuWTCent VOll........................................................................ Voice Over Internet Protocol SC .............................................................................. Technk~i! Support Cenzer.................................................................................... u................. Wr u A-4 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 (NOUE)*°: 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 NEl 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 descending 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.Initiating Condition (IC): An event or condition that aligns with the definition of one of the four B-I 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: The barrierts) between areas containing radioactive substauees and the environrnent.,(Inse.t a site sp.... definitian fcr this- te-"m..)-r .. .. C(i)NI'AINM LN I IN II (RVITY: Primnary (Containmentn (}PL.RAlIlI F per Technical Specification 3.6.1 I.. Secondary ('ontainmntn OPI ,RAI}I l; Fpr 1 echnical Specitication 3.64.1I 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.: fAUTE:fhftrmapledt alc stea geeaa ht p ta ek the .....ar side of suffcient sizec cau.....e uncantrae dra in steam ... gernerater pressure er th steam.PWP~s 8~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 power 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 (UC(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. 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.B-2 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.t;:enty, four hourc cccluding the c.:'eurt peak value.OWNER CONTROLLED AREA (()CA): The site property owneod by or oiher'.x ise uinder the lice..... In ........ "ac it .ma b appropri.ate for a liene t... dein a++ .m..... are .... th a perimeter closer tote plant Protcte perim.. (.g. a; ..t .with a l.r.. OCA, ..h. ...m..PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.PROTECTED AREA (PA): The area thai cncoinpasscs all controlled areas the protected area fenee. (Ince," a site epecifie definition for thi:; trem+.) De'.elop.er N:ote Thic term ic t3Tpeal!:,ly ta ..en.to mea th .... are und ..r :otnuu ......... m .nio.... an contr.l an arme .d protectio

o- dec..... d the Soecurity Plan.REFUELING PATHWAY: l ift include> the reactor cavity, the transter canal, and the spent fuiel rbool.(l:ncen a .c:*e 5Fcpeeiti b-r t+" t..... ' ev.lo'p-r thic+ Jc.....o ch.....not ;.includin the, reator ,,cosel.

The..condition of a eteam gen'erator in whkich primary to ;ccondaory lea~kage of"+ .......n ma ,,nitude ..to re -ir a in'jection. De-veloper Noet. -Thi: t..rm i 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. Develaper Not: Thic te'rm, may be mo~dified to include thec a~ributec cf "cafcty, related"' in aecorda'ce wi;'th !0 CFR. 50.2 or other cite cpeeifie te.-rmino..logy., 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.UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED: A paranmeter change or an event that is not 1) 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.B-3 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 SG1 ECL: General Emergency Initiating Condition: Prolonged loss of all offsite and all onsite AC power to eamergencey\>ssntiuI buses.Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director shot,4d-x ill declare the General Emergency promptly upon determining that sp..ee"I4e hours-) has been exceeded, or will likely be exceeded.(1) a. Loss of ALL offsite and ALL onsite AC power to 160 X/AC en¢tla Buses AND b. EITHER of the following:

• Restoration of at least one AC effirgencycsscntial bus in less than .i-tpee4iqe4 hours-) is not likely.* Recactor xessI x ,,aicr k".cl cannot bc restored aud m~aintineiud athuxc Mdinin um Basis: This IC addresses a prolonged loss of all power sources to AC emeegeineycsscntiaI 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 sh \il-i Il require declaration of a General Emergency prior to meeting the thresholds for IC FGI1. 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 etmefgent) essential 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 emefgeneycs~scnmial bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should 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.131 Aithheugh thie ,C an....... a...... iew.d....r.d.nd....t.. th.............d.....arr..r....,..t ic included ........i:e.fr a.. more timely of the ern.ergeney ,laci.cat:n. , l....l.T....... t-t- ifie encrgte. y...... '...... h......... f.d......... t... r.. m.. rg. n.y......... .......ic typically, .... ..m.rg.n.y b.,u.cL., per train ef SAFETY SYSTEMS.The "citec pecifie houre" te rcctore AC pew;er to a em emrgency buc cheuld be ba~cd cn the Gui'-de 1. 155, St~rioh BIhwckoa:t. Site 3cpcific irndicaticn of an inabfilit te adequately remove: heat from the co-re:........ Rece .ec.....J ..r...;'"' I~dl ,nof b r.to-ed andl maintai.. d ab ......... Miiu S.. am re :a-er thel atep.r le,,v.l tha..ri... ntry.nt..a..r....ingr ......i.n.....ur...r..thrwic reqir .iplamntatiuon ..f _prp gc.i..... Altrnt..l.. a... citeg may uc ........r I A tAiL Acci~nment /'dtributec: I.t.lJi 132 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 sheo4d-ax ill declare the General Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) a. Loss of ALL offsite and ALL onsite AC power to 16 VAE ti 1 1, L.F A specific emergency buzez.) for 15 minutes or longer.AND b. Indicated voltage is less than 05/210 VI ste specific bus -veltage ;alue) on ALL 2/ I 2R2S0l6~J 212-S0O 7(sitc specific Vital DC basses) for 15 minutes or longer.\kU 4 ~AVAt~ 4~b~a~ioz~ 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.Ffenminutes was ...l...t.d as a the threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.133 Th~ 'zitz cifi~ Vital DC bu~z" arc thc DC bu~c~ that pr~;idc rncnitcring a~d zcntrcl capabiliti.z~ for SAFETY SYSTEMS.This IC and EAL wara addad tc Rcvi~izn 6 ta addrca~ zpcrating c~pzrizncc fram tha March, 2011 aczidcnt at Fukuahima Dpi jahi.ECL Aaaignmcnt Attributca: 3.1.1.B 134 SS1 ECL: Site Area Emergency Initiating Condition: Loss of all offsite and all onsite AC power to e ergenecyssenial buses for 15 minutes or longer.Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: SNote: The emergency director §heoltd-\x ill declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Loss of ALL offsite and ALL onsite AC power to 1.VC i I ZE.i/2F o sit specific ..m.rg.n.y bu....) 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 wes-se eeted-as is a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level wo4-v.,:be -*iauscs ICs RGI, FG1 or SG1.Del~epe~s-rNe.'eM Far a pev.'r source that multiple generators, the EAL and'er Basis scti:en she.'uld reflect the i.e.,.t. 50% capacity generators sized ta feed I AC emergency bus), the EA an, Basis: The "site specific emergency busesz" are the buses fed by effs ite or emergency AC pow:.er sources that supply pewe'r to the electrical distrib.tion .......m that, p....r. SAFEctTY SYSTEMS. is typically I emergenc.. bu per.. trai ef SAFETY SYSTEMS The.EAL.and..r..asi.....t.. nmay u............. saet rel.ted pe ........ prvid....th-at operation of this sc~urce is contrelled.- in .. ..._.aee .... with abnor.mal ar emergency ..p.. rating:-proc.edures.. er beyand design basis acci:dent resense guidelines (e -g., FLE ......... guideies)At multi unit statons the may credit compensatory measur"es that are proceduralied a.nd pr..d.. s etc.... .Plnt ...at have .. a proceduralized te .. suppl ^offsit AC power te an 1354t 60 rn i

• .oA f affe~ kd unit ;Ia a zrZ ~z~zcmr~anxcn unit n~ay ~r~Iit tflIa pzw~r ~curzz in tfl~ L~L:m;'idzd that th.~ vipnncj Crczz ti~ ~trat~zv mcct~ th~ rc~iuzrzzncnt~

CI IU Li-K YJ.t3.ECL ,Azg nmzrnt ,Attributec: 3.!.2.136 SS5 ECL: Site Area Emergency Initiating Condition: Inability to shutdown the reactor causing a challenge to RPV water level or RCS heat removal.Ope rating Mode Applicability: Power Operation Emergency Action Levels: (!) a. An automatic or manual scram did not shutdown the reactor.AND b. All manual actions to shutdown the reactor have been unsuccessful. AND c. EITHER of the following conditions exist: ai~xjuta!y hemt frcm ccre) ~w Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram 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 tlht~s-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 in that 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 weteld-be-v4*usc-, tC RG1 or FGI.137 De~epe*-Nete~t L I t I fliC Ri iC applieatle in any MaCic in wfliefl tne actual reactar p..,~.. L. ,A .,.,uia exeeco tne.. o! + q I ,I....... I...... Ie'c at.. , h+.c..fl tf. rcac.tor iC caCijeredC huJtd.?;vn.AI". A with .a,*. ,.hutda;.wn rcactar.p....r lee th. t i :cc tha r ....qual ta the. r-eactar paw;;er level whic:h dAfec thc_ .......... d.PwrOprata (Ma"d- I) wil" n... to include. .ta.tu. (.ad 2) in the Opc..t.. :-_ de Aplc;-abiiy.Fratcn, mp..,.if.....r...t.r............- t, ..... zhtn ..t. 3% a.... P.....Operatian-sc'tsz ..t t..n h.e ,C pisals applicable in Sta."tup an EA.L etatement, the.. Bacic: or bath ,-.g.,t" a reactor power level).Site Cpacific indicatian afar. inability Ia adequately remav'e heat fra~m the care:[P;U,. Inpertn ;CJite+ ....fi ., far an nae/uarey reoie thenaauple temperature anoi'requi,,rJeC itperHetatia Caprampt Tectratiurn acmtia. lTeisnatdelysasiete mnayi~;'t ucm'e iae/art ,exit thA niacatpe great.ertha Ia:'e ,20F anda ratr ee.wae evl FCL .Meianment AttributeC: 3.l.3.B 138 SS8 ECL: Site Area Emergency Initiating Condition: Loss of all V4qal-x ital DC power for 15 minutes or longer.Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: SNote: The emergency director sotidd-x~ ill declare the Site Area Emergency promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) Indicated voltage is less than VI sp~cific bus '.ocltage value) on ALL 125i7 VD us1222$I6 fR22 p....i Vi...........l.: DC,. b~-uz.... for 15 minutes or longer.Basis: This IC addresses a loss of vital DC power which 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.Fifteen minutes wis eiiN a threshold to exclude transient or momentary power losses.Escalation of the emergency classification level we.*.uldbe ICs RGI, FGI or SG8. Tha %!te s~pcifk vezltagz value" h be "de zn tha minimum b'-c ;'eltage ncc.......... ............ e-...:' fS.AFETY SySTEM equipmcnt. -Thic...o rtagz ;value chould ince.'boratc a ma.rgin ef at leact 15 m~nutec of z.pemticn hafore- the .....t cf inailty:;. t........thc.e load.;..! Thi " vc ........ ucuall ner th ..minimum ~'la rc. J .. h.n-,,tk c..Izi ezntrc,! capabilitices fcr SA*FETY SYSTEMS.ECIL Acc!gnmrcnt ,Attrib!ut.ce: 3.!.3.B 139 SAl ECL: Alert Initiating Condition: Loss of all but one AC power source to emsergencevessential buses for 15 minutes or longer.Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director declare the Alert promptly upon determining that 15[minutes has been exceeded, or will likely be exceeded.(I) a. AC power capability to 1 A i alD /EIF: .O~~e spedific emergency bu~ae) is reduced to a single power source for 15 minutes or longer.AND b. Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS.[73k: Vt~ 4 6Oy~4C P~ses tntqp~ti~I L741VI34I~VACBu~r~tio~i Basis: SA -12k IV y~r SY ,l F: A s?,'stertn required for salte planlt ope~ration., cooling down the plalnt anid/or placing ii fin the cold shutdownl condition. including the LC(iS. These arc typically s~stemns elassitied as siifet y-related. This IC describes a significant degradation of offsite and onsite AC power sources (see '1 able S 1 ) such that 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 SU 1.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergececycssenI ial bus. Some examples of this condition are presented below.* A loss of all offsite power with a concurrent failure of all but one eimiegeineyessential power source (e.g., an onsite diesel generator).

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

140

• A loss ofe frefiey.>scntial power sources (e.g., onsite diesel generators) with a single train ofeffiegenecysscnfial buses being back-fed from an offsite power source.Fifteen minutes we-eeee-is a threshold to exclude transient or momentary losses of power.Escalation of the emergency classification level -we~uld-bc viuscs IC SSl1.minimu numb.. r of_ eperteing gen.... r,. n.ce....ary for .th.........eto.pr..ide required power to an AC..m.rgency hu,.. Fo ........ if fa backup p....r.......

ic ......ce ;of ftw- generator.o (it:e..,," no/ 5 .... caact ........c sized. toA fee ,AC-b'us), the* [,AL .ad Basic must'c sprcify that.. bothFc gnrtoree fndr that sorc ar5e oerating.pwe rc! (e.Them rc;ite sp fe st pcificeere cybs" UFSrPe the buse :,sife by of~chitd or: emfrency Cr powz orer.... r hud oiy h ulee sape ro..id..d in. rhebasid se:e cti-n, abo e;*a nded The E.Lc and'e Bosic :chtud eflc that pcoch indepen.n....t....r....t ontiut tha perati..n

- f fthic source ic recognized in A.OPc rand EOPs, or b}eynd decign basis ..c. dentmulti unit" tations.'

the , Eac 3 credit co natoy mredit ur th..t...........d:ralized A n provldeJ that the piannel crocc tic ctrategy meetc the requirements ci U ~K ~ECL Assi~mnent Attribotcc:

3.1.2.8 141 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: Note: The emergency director she*4d-w ill 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 1 5 minutes or longer.Reactor Power RPV Water Level RPV Pressure Primary Containment Pressure-Suppression Pool Level Suppression Pool Temperature 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 scram* ECCS-{S-) actuation* Thermal power oscillations greater than 10%Basis: l NPI ANNI';!): A parameter change or an evenot that is not 1 ) the result of an intended es olution or 2) an expected plant response to a transient. [he cause of ihe parameter change or evenA may he known~ or ulnkn,,)\,,n. 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 thus represents a potential substantial degradation in the level of plantI safety-of-te 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.142 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 would be 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 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, RiPV level 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 RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.Fifteen minutes. .... ......et...d a o ais the threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level .....itA eus ... s Fs loICR1 In the. P aamecte~r list c~olumn, the.. ..it. scii ... mb..r...........r........t....mini..um num:ber of steam gen ~rat r nc ary-' for plant cooldown; and "shutdown. This crite~rion may also speei!.*' whether lee ... lue .,. o ... be.wid rang,.na.w rng..........

r. b.eth, dpnding up...n th monitorin'g require~ments in emergency epe~rating procedures, De...l.pers may. spwcify ..ither. pres.uri..r.r.....~....

e.... I. in t.e PWRo paramete~r co'-umn The type-, location layout of Control Room indic~atio~ns, and the r... of possible faiur m.d... an chall. nge the.. abiit of:" o" operator to ac~curactly de~termine, within the time period available for emergenc~y classification assessme~nts, i.f a specific perc.entage of indicatioens classification: asessme~nts by focusing on the in:dic~ations for a sel!ected subset of parameters. values, the EAL recoegnizes and accommoda.ctes the wi;'de -variety of i.ndications in nuclea=r power plat Cnrol R... oms... Indiation: soeurcecs ma~c:.' be analog or digital. saft r...e+ate or nopfmi orater -ate,idvdal.mete value or com.pute~r groupdspa.ec A l.o!ss of pla~nt ann:unciators will.. ..* ,.) .. ....e.aluatd for reportabiliry' in accrdan:ce CR, to cerfcrm emero'enc\' assessments. Comp-ensatorv m'.easures fcr a less ef 143 a.nn.unciatienz radily, implemented anzd may in.iudc inereezed menitering ef main centrel.............. r .frqun ..... rend b...... nen ........... eratere. Their alerting fu:netien infocrmatie c .:ed te eprate the plant. er ureeec threzugh AOPe er [OPe. Bae~zd en t.he.c requirercnent of 10 CFR. 50.72 (and .....te guidan.e. e .n.. EG10. ; lllThe repo.. in. ".fthiel.appropriate compeneatery meacures and cerrectire ati:one. ......diti..., aae ..f............ thei:r cite epecifle EAIs.-h refueling, and defueled medee. ne ar.alagoue IC ic included tbr thece mede: efoperatien. EUL Aecienment Attrlbutee: 3.1!2.B 144 SA5 ECL: Alert Initiating Condition: Automatic or manual scram 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 Emergency Action Levels: 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.(I) a. An automatic or manual scram 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 scram 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 plaint safety-ef the pla~n. 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, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). 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". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action.The plant response to the failure of an automatic or manual reactor scram 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 RPV water level 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 145 possible via IC FSI. Absent the plant conditions needed to meet either IC SS5 or FSI, an Alert declaration is appropriate for this event.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.Thic IC ic 3pplica.ble in any, in w:hich the a.ctual". rea.cter pew:,er level ceuld exceed Opentien ....... at / ,5, then teiC ic appl:ca.ble in Startup. Made ECL Accignmentn Attributec: 2.1!.2.8 146 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: H"R!: (iombustion characterized hy heat and light. Sources ot smoke such ats slipping drive helts or overheated electrical equ~ipment do not constitute FIRES. Obhservation of flame is preferred hut is NO I required if" large quantities of smoke and heat are observed.FXPI()SION:

A rapid, violent and catastrophic htailure of a piece of equipment due to combustion, chemical reaction or overpressurization A release ot steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing.etc.) should not auttomaticallv he. considered an explosion. Such events may require a post-event inspection to determine if the attributes of an explosion are present.SAIFE ]Y SYS ['EM: A system required for plant operation. cooling down the planmt and/or placing it in the cold shutdown condition, including the I-CC'S, [hese are typically systems classified as safety-related. VISIBLE DAMAGE: Damage to a component or structure that is readily observable wxithout measurements, testing, or analysis. fhe visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure. 147 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[Al. l.a identities haiardous events that could result in damage to plant sy stemas. A seismic c,,cft is inldicalted bS entrV into IC 11,12. [looding is indicated by a significant increase in \xatcr lev els (external or internal). Ilich arc indicated by sustained v, inds at the site meteorological toxsor exceeding 35 mph.The tirst thresho~ld fbr EAL l.b4- 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 should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.[he second threshold tbr EAL l.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 the totality of 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 weald-be-i ,,:e IC...o.RI ECL A~ignment Attributes: 3.l.2.B 148 SU1 ECL: Notification of Unusual Event Initiating Condition: Loss of all offsite AC power capability to ewiergenceye.ssenial buses for 15 minutes or longer.Operating Mode Applicability: Power Operation, Startup, Hot Standby, Hot Shutdown Emergency Action Levels: Note: The emergency director ill declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(i) Loss of ALL offsite AC power capability to Esv 1 s I/2I! 112 i2 f l..it .. sp ....r..nc. btss r 15 minutes or longer.i 1t~ jD IR 2 C7~1: V~ ~ 416Q vAE ~p% Wb~ou~Basis: This IC addresses a prolonged loss of offsite power (see 'labic S2 ahux c. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emer'genec;'ssentia! buses. This condition represents a potential reduction in the level of phcrn safety-e4f-41e plan4.For emergency classification purposes, "capability" means that an offsite AC power source(s) is available to the emergeneycsscntial buses, whether or not the buses are powered from it.Fifteen minutes wa slete a...... ak ihc threshold to exclude transient or momentary losses of offsite power.Escalation of the emergency classification level w;e',dd be-v4auscs IC SAl.The ..i.. sp.cifi. emer....nc.. bues .... t*. he. bue fe.. byA or ...urc..is typically I ..m.rg.n.y per train of SAFETY SYSTEMS At mult unit stations, the EALs may credit compensatory m.-:easures that are procedu-ralized and procedures. etc. Plants that have a proceduralized capabilib" to supply cffsite AC power to an affected unit via a cross tie to a companion unit may credit this pewer source in the EAL prov-ided that the cross tie ......~ met the..... .gire.......f..0..... 5.63....149 ECL Assi.nmcnt Attrib"ut: 3.!.I.A 150 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 shouild-xx ill 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 RPV Water Level RPV Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature Basis:[ NPIANLI)D: A parameter change or an exent that is not 1 thea result of an intended evoluittion or 2) an expected plant response to a transtent, Ihe cause of the parameter change or event ma.be knlowni or 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 safety-ef the plaft.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.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 ,etid4-be-is reported if it significantly i mdi pairs the capability to perform emergency assessments.=- t-nparticularly those necessaty to ;amcrgency acazzmz~ta necezsary ta 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, RPV level 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 151 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 RPV water 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.Ffenminutes w;,-selee ..... si a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level ":e*-Id bc .'4aiscs IC SA2.In the PRWP parameter lis ....u.... th .... te* specific number" s!hculd. retlect the minimum.menitzring requirements in emergency operating procedures. Th....um...r, *ypr ............. a ....u .....C........................ a, th.... range ecfpible fe uemodes, ca ..hall..ng. th ..ility of.- an pc-retorr.---' t...c.u.ately dctcmin: .ithin th. e time: period available for emergency classification assessments, i:f a specific percentage Cf indications have been lost. The app--aech used in this EAL facilitates prompt and acc"urate emergency elassificatian "oss....sments ..y fousn ..n th'e indications for" ao od :ubest of parameLrs--. By. focusing on the- av'ailability, of the ..p.cified-parmr.eter .,alu...es ... intadfth so~u......rce .f.h..plant Con... ,tro .... Indiaticn ....... an ........ may be ana'-g er digital. safety, relte ...oer A less of plant annu-nciators -w.ill b ev'alu.ated fcr repc:rtability' in accordance with 10 CFR 50.72 (oan the .......c.iat guida"nce in NUREG 1022). and repCrtced if it si'nfi:nt' the capability/te perfolem ....ergen.cy o.ass'ssments. CompensateD' measures fcr a Ies5 cf b..rds ;an more. fr... quent. pl" n ....... by:: .... licend 'erator. Thi nC.........d ng .............. dc not..... prvd the... prameter............p.ci.. cmpCn._enttau crniteriaand athaefr nct included,.o ... i... thiusr IC an meresgL. cn ee 152 thzir site zpafc!q~ EAL:.Due tc. chmagc: in thza ccnftguamticr." o~f SAFETY SYSTEMS, inaluding az~cciatad

i...........

a ... and indi;..izr^_. during thc ec!d rcfuezling. and-q dzfu.kd m^dz, na................ Iv U--k., L, / i "gU. / ",,;,;.r[ : &, i. ;.,", 153 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: (I or 2)INote: I Isc the I T nit I or I lnit 2 lPrctrcatrncnt vs. ntRihr) (iraphs to deterinen if thc Pretreatment Radiation Monitor cxcceds the TV o[ 24t).000~ pCiiscc.(2) Sample analysis indicates that reactor coolants sFcitic activity ,ahte-is E! I IIER:* r*tkgink ~ q~ ltif doe lt3 for* raethn.1 e/adoeettia ntnireacrha .al ,al l.mi .ncfe I 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 efthe-plant. Escalation of the emergency classification level weu!d-be-v4aoses ICs FAI or the Recognition Category R ICs.er EA han Ie! En~ter the radiation mcnittrc) thatamaynge--*ith tre adil-ytc idntifryar -wh 'en RC aciiylvl xedTechnica! Specification allowable limit. Ti A a.b eeee!f newre capnbiexitieicng t rzte, qu~aired) E fmpeeo detiet.ing thmEAthhed n paiitic inclu:dne: e!zd 154 F~r EALit2 D~vJ~p~rz may r~wcrd thc EAL tc incluic th~ rcactcr ccalant activity parwnctcr(~) 5pcciflcd in Tcchnical Spcciflcatiznz ~d thc ~cciateJ allzwabk Iimit(~) (c.g., valucz fcr d~zz zguivalent I 131 and grc~z activity, zinc dcpcr.dcnt cr flansicrA va1uz~, ctc.). H thL appr3ach i~ ~clcctcd, all RCS activity allo;vcblc Iimitz &h3uld bc includcd.EC!. .X~i~nmcnt Attributcz: 3.l.1.A and 3.1.I.B 155 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)SNote: The emergency director sheokld-w ill declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.(1) RCS unidentified or pressure boundary leakage greater than gpn 7!3 VJ ~44S ~ ua for ! 5 minutes or longer.(2) RCS identified leakage greater than ,'ite z~pcitic v'alue ,5gp or 15 minutes or longer. jy :r4 (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 which may 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 plant safety-ef the-phsii. 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 UNISOLAIBLE 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. A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.Escalation of the emergency classification level wouldbe v4auses ICs of Recognition Category R or F.156 in th: ;it&': T~chni[ca1 Spzf. tbr this .t~p zf [z-ka;=z.E AL f-, F..or. th : ...... :pcifi '"a' ..... -'a .enter thc highc f2 p rthtu pof~in sitc'h Thzhnica! Spczifiz:.aticn fcr this of for IS zr !Ing~r.157 SU5 ECL: Notification of Unusual Event Initiating Condition: Automatic or manual scram 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 scram 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 scram 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 scram 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 scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram 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-4f4he-plea'..

Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). 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 scram 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 scram using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram 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.158 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 scram). 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". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action,.The plant response to the failure of an automatic or manual reactor scram 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, and other concurrent plant conditions,-ete. 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 FA I. Absent the plant conditions needed to meet either IC SA5 or FAI, 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 scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance be applied.* If the signal causes a plant transient that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, and e*dxx ill be evaluated.

• If the signal does not cause a plant transient and the scram 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.-Th:... IC" is applca.ble in n:. ...d..i...hi.h...h........

reactar paewer level c..u'd exceed the..paw"er !-v;cl a.t ,which thc reneta~r ic caencidered chu-tdon-. A PWR --ith reactor anEL-atmnth.ai ........... bt (.g..., .. rece poe rcFcc ;h lvel)."Fr~!t

...L .~-.ccignmern

.'~.nrit~i.aec: .c. 1.1 159 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: (1 or 2 or 3)(1) Loss of ALL of the following onsite communication methods: (2) Loss of ALL of the following ORO communications methods: FNN (Frnergency+ Notification Netx, ork)( onnuercial phones (3) Loss of ALL of the following NRC communications methods: L NS on Federal Telephone S;} sten, FI) (onmmcercal pho nes (zi.te spcci.ie l:st zfcc....... zn methc.... d.,,.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 should be assessed only when extraordinary means are being utilized 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,--~et). 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 thle state of (ieor~ia. Appling County.J lif lDa\ is Connt',., l'aunall Counlty and T oombs Count" '4+ r"... .. .. D vecz .os...x...+. EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. 160 EAL #! Thc~ "shtc spccifc !ist cfccrnmunic-"ticns mcthcds" shsu-ld includc all..................... 1...... d.. .. .......... nt ..... ......... ......s :cg., cs.nm crcial cr sit: tc-p lc s.v ............... c ....... s,......i..l.. tin......lm ine.' c ld..i....... pl..EAL fl2 .The "sit specific list af methads" shauld! incluadc all oom.'municaticn' "',.-Em........ ,1-hc.. listin';*;g sheuld include, ir~stall.d pl a.nt cguiprncnt and ccmpncnt, s, and net^iterns awnad an maintalned by, Era-mpl: metheds rin.g dw-:,'d:dicoated cemmunicatiens techneelegy. typically" within= 15 minute;.Emergency The li.tin. shul ....l...............d pl..t cquiprne-nt an.... ce .......... an..n.t items e.w-ned a.nd iaintained b) These m'ethods tpically' the dedicated... ........................... .l [Il.C 161 APPENDIX A -ACRONYMS AND ABBREVIATIONS AC .................................................................................... Alternating Current AOP ..................................................................... Abnormal Operating Procedure ATWS.....................................................................Aniiae Tvransen PWitou S~n c AW ................................................................................ ou BScra&W............................................................................. Babcockc and Wilcc'................................................................................................. B liT Bcc .... ticn_,_ !nitiatian Tc.m.pcraturc lBl l)(G.............................. .................. Building, BWR .............................................................................. Boiling Water Reactor (1B .............................................................. C onirol 1B uilding CC.......... .......................................................................... uhie Centimeter CDE......................................................................... Committed Dose Equivalent CFR......................................................................... Code of Federal Regulations ................................................................................................ C SFS Criti"a Safety Functia .aw Ic................................................................................... Cint .... cidr~(P .................................................................................. (....u......P.r..Minu...e (SFT.........................................................................(Couc!SfeyF ncts er Second re..................................................................................... Direct.....Current.. BA 1................................................................................. Doseg Iqialent Iodiet (1 ........................................................................................ D nl r Minuel EAL................................................................................Emergecy Actioalnt Levele ECCS RM..............................................................Emergncyl CodeRae Cooling Syste ECL .......................................................................... Emrec ClassiencyAction Level ECS....................................................................lEmergency NotifCotiing Sytstem EL 2...................................................................... Emergency Noassification Levtel EPA.............................................................Environmental.........Protection......Ag .Enc............ PG...............................................................Emerency, Proedue .Gidein PIP ................................................................. Em Irg c rgcy Pl Notification Networkr PNR ...................................................................... Emcrutinar Povifcazr SRscacz EPA ............................................... Envctiro m na Pcw rotection Agncyiut ERG..................................................................EEegnyPa mpcrgenc'y: Prc~cpodu G AXviation Administration A-1 t:B ...................... 1.............................lederal lBureaui of Investigation FEMA ....................................................... Federal Emergency Management Agency FSAR....................................................................... Final Safety Analysis Report Il ..................................................................... Federa~l l'elecoffmunllieationls Systemt GE .................................................................................... General Emergency iM ....................................................................................................... (iram HCTL................................................................. Heat Capacity Temperature Limit INI P.....................................................................................I latch N uc ear Plant HO ................................................................ Hlcadquartcrs Operations Officer (NRC)HPCI ................................................................... High Pressure Coolant Injection................................................................................................ H S! Hum..n Sy':tam Interface IC ......................................................................................... Initiating Condition................................................................................................. ir PEF.......Indi,,idu.. Plant Eam..t.. af,: Ext ..rn..l E-ventsq (G.enearic Lette,;r 20)ISFSI...................................................... Independent Spent Fuel Storage Installation ................................. ............................................................... K................................................................................... n ...................... Fa t CO...........................................................................Liitn Canditian if Opemtien LOCA .......................................................................... Loss of Coolant Accident................................................................................................. CR.................................................................................. Mi a'r......N MSL ..................................................................................... Main Steam Line t i.......................................................................................n m cro-Cturic mR, moRero, toremo, mREM ........................................... milli-Roentgen Equivalent Man W ............................................................................................... g.-Nat NI!"................................................................................................ Northeasi NEI............................................................................... Nuclear Energy Institute NPP.................................................................................. Nuclear Power Plant NRC ................................................................... Nuclear Regulatory Commission ................................................................................................ N NORAD............................................... North American Aerospace Defense Command (NO)UE ................................................................ (Notification Of) Unusual Event............. .................... ............................................................... N..................................... Nuclear Nlanagcmcnt, and Rcacurcce Ceuncil OBE.......... ................................................................ Operating Basis Earthquake OCA.............................................................................. Owner Controlled Area ODCM ........................................ Offsite Dose Calculation (ssesm........-Manual ORO..................................................................... Off-site Response Organization PA .......................................................................................... Protected Area A-2 ACS................................................................ Prierity ...Actu.....t ...... .. an Cnr,... cal ..... ....PAG.......................................................................... Protective Action Guideline I1X ................................................................................ PIrivate Branch Fixchange P S.................................................................................Pn Srstc ICS .......................................... Przceca and Ce~ntral Syctam PRA/PSA .......................... Probabilistic Risk Assessment!/ Probabilistic Safety Assessment ................................................................................................ P WR............................................................................. Prcc:'-rized Water Reactor................................................................................................ P S ................................................................................. Prtection Syatem PSIG...................................................................... Pounds per Square Inch Gauge R................................................................................................... Roentgen RCIC .................................................................... Reactor Core isolation Cooling RCS .............................................................................. Reactor Coolant System Remn, remn, REM.............................................................. Roentgen Equivalent Man RI1R .............................................................................. Residual I leat Rcmo\ al RPS ........................................................................... Reactor Protection System RPV .............................................................................. Reactor Pressure Vessel VLIS....................................................P.Reeactr V'esce! LeveI !netrumentatian Systemn RWCU............................................................................ Reactor Water Cleanup Rx ...................................................................................................... Reactor SA(U..........................................................................

......eer Accident (inidel inc SAR...............................................................................

Safety Analysis Report A ................................................................................ ...A S ............................................................. a........ itr an S!actern'N.............................................................................Secondry..o.....Sn SCA......................................................................Stl-Cnane ratling Apparautu................................................................................. nd... r......nt....nmen..t GSC.................................A................................................Sl-otie Breathn Apparatusr ................................................................................................ S S P.......................................................................................Sta Senaratar SF ....................................................................................... SlPoo SC......................................................................................ompan SPS ................................................................................ SaeyPretrDslySyuheste .................................................................................................. r......r SW\ .................................................................................................. South west A-3 TEDE................................................................... Total Effective Dose Equivalent TOAF.................................................................................. Top of Active Fuel IV ........................................................................................ lhreshold Value VAC' .......................................................................... Volts Alternating Ctua'nln l)irecct CuWTCent VOll........................................................................ Voice Over Internet Protocol SC .............................................................................. Technk~i! Support Cenzer.................................................................................... u................. Wr u A-4 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 (NOUE)*°: 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 NEl 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 descending 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.Initiating Condition (IC): An event or condition that aligns with the definition of one of the four B-I 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: The barrierts) between areas containing radioactive substauees and the environrnent.,(Inse.t a site sp.... definitian fcr this- te-"m..)-r .. .. C(i)NI'AINM LN I IN II (RVITY: Primnary (Containmentn (}PL.RAlIlI F per Technical Specification 3.6.1 I.. Secondary ('ontainmntn OPI ,RAI}I l; Fpr 1 echnical Specitication 3.64.1I 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.: fAUTE:fhftrmapledt alc stea geeaa ht p ta ek the .....ar side of suffcient sizec cau.....e uncantrae dra in steam ... gernerater pressure er th steam.PWP~s 8~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 power 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 (UC(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. 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.B-2 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.t;:enty, four hourc cccluding the c.:'eurt peak value.OWNER CONTROLLED AREA (()CA): The site property owneod by or oiher'.x ise uinder the lice..... In ........ "ac it .ma b appropri.ate for a liene t... dein a++ .m..... are .... th a perimeter closer tote plant Protcte perim.. (.g. a; ..t .with a l.r.. OCA, ..h. ...m..PROJECTILE: An object directed toward a NPP that could cause concern for its continued operability, reliability, or personnel safety.PROTECTED AREA (PA): The area thai cncoinpasscs all controlled areas the protected area fenee. (Ince," a site epecifie definition for thi:; trem+.) De'.elop.er N:ote Thic term ic t3Tpeal!:,ly ta ..en.to mea th .... are und ..r :otnuu ......... m .nio.... an contr.l an arme .d protectio

o- dec..... d the Soecurity Plan.REFUELING PATHWAY: l ift include> the reactor cavity, the transter canal, and the spent fuiel rbool.(l:ncen a .c:*e 5Fcpeeiti b-r t+" t..... ' ev.lo'p-r thic+ Jc.....o ch.....not ;.includin the, reator ,,cosel.

The..condition of a eteam gen'erator in whkich primary to ;ccondaory lea~kage of"+ .......n ma ,,nitude ..to re -ir a in'jection. De-veloper Noet. -Thi: t..rm i 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. Develaper Not: Thic te'rm, may be mo~dified to include thec a~ributec cf "cafcty, related"' in aecorda'ce wi;'th !0 CFR. 50.2 or other cite cpeeifie te.-rmino..logy., 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.UNISOLABLE: An open or breached system line that cannot be isolated, remotely or locally.UNPLANNED: A paranmeter change or an event that is not 1) 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.B-3 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}}