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Attachment 2 - Markup of Seabrook Station Emergency Action Levels -Initiating Conditions, Threshold Values and Basis
	ML16068A131
Person / Time
Site:	Seabrook
Issue date:	02/27/2016
From:	; NextEra Energy Seabrook
To:	; Office of Nuclear Reactor Regulation
Shared Package
ML16068A128	List: ML16068A130; ML16068A131; ML16068A137; ML16068A139; SBK-L-15120, Independent Spent Fuel Storage Installation (ISFSI) Ics/Eals;
References
	SBK-L-15120
	Download: ML16068A131 (67)
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{{#Wiki_filter:ATTACHMENT 2 Markup of Seabrook Station Emergency Action Levels -Initiating Conditions, Threshold Values and Basis SEABROOK STATION EMERGENCY ACTION LEVELS INITIATING CONDITIONS, THRESHOLD VALUES AND BASIS TABLE OF CONTENTS 1 REGULATORY BACKGROUND ..............................................................

1.1 OPERATING

REACTORS ........................................................................... 1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI).......................... 1 1.3 NRC ORDER EA-12-051 .................................................................... 2 1.4 ORGANIZATION AND) PRESENTATION OF INFORMATION................................. 12 1.5 IC AND EAL MODE APPLICABILITY........................................................ 13 2 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS............................ 21 2.1 GENERAL CONSIDERATIONS ................................................................. 21 2.2 CLASSIFICATION METHODOLOGY ............................................................. 22 2.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS .............................. 22 2.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION ...................... 22 2.5 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING ............ 23 2.6 CLASSIFICATION OF SHORT-L1YED EVENTS............................................... 24 2.7 CLASSIFICATION OF TRANSIENT CONDITIONS .............................................. 24 2.8 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION........25 3 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS .................. 26 4 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS ............... 42 5 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS .......... 65 6 FISSION PRODUCT BARRIER ICS/EALS ................................................ 67 7 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ......51 S SYSTEM MALFUNCTION ICS/EALS..................................................... 104 APPENDIX A -ACRONYMS AND ABBREVIATIONS ........................................... 1 APPENDIX B -DEFINITIONS..................................................................... 4 ii

OF EMERGENCY ACTION LEVELS 1 REGULATORY BACKGROUND

1.1 OPERATING

REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities. Several of these regulations govern various aspects of an emergency classification scheme. A review of the relevant sections listed below will aid the reader in understanding the key terminology provided in Section 3.0 of this document.* 10 CFR § 50.47(a)(1)(i)

10 CFR § 50.47(b)(4)
10 CFR § 50.54(q)* 10 CFR § 50.72(a)* 10 CFR § 50, Appendix E, IV.B, Assessment Actions* 10 CFR § 50, Appendix E, IV.C, Activation of Emergency Organization The above regulations are supplemented by various regulatory guidance documents.

Three documents of particular relevance to NEI 99-01 are:* NUREG-O654iFEMA-REP-lI, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]* NUREG- 1022, Event Reporting Guidelines JO CFR § 50. 72 and § 50. 73* Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors Emergency Preparedness staff.1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)Selected guidance in NEI 99-01 is applicable to licensees electing to use their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR § 50 and the guidance in NUREG 0654!FEMA-REP-1. The initiating conditions germane to a 10 CFR § 72.32 emergency plan (as described in NUREG-1567) are subsumed within the classification scheme for a 10 CFR § 50.47 emergency plan.The generic ICs and EALs for an ISFSI are presented in Section g5, ISFSI ICs/EALs. IC E-44U-1E U I covers the spectrum of credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to !.inslations. or facilities tha my pocss n4or epckae pen fie (eg. a sntord et~vale toag The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-l1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees. NUJREG-1 140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 remn Effective Dose Equivalent. re~uirea Icr IU t.A-K ~ ~ emergency pian Is ornerent man mat presermep ror a mu CFR ~ 50.17 emeraenc'i plan (e.a.. no emergency technical support funetien). 1.3 NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2012, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety fractions of core cooling and containment integrity, and ultimately led to core damage in three reactors. While the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling.Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii). Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event, To this end, the NRC issued Order EA- 12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license.NRC Order EA-1 2-051 states, in part, "All licensees ... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." To this end, all licensees must provide:* A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;2

A display in an area accessible following a severe event; and* Independent electrical power to each instrument channel and provide an alternate remote power connection capability.

NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify~Licenses with Regard to Reliable Spent Fuel Pool Instrumentation ", provides guidance for complying with NRC Order EA- 12-051.NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order BA- 12-051. These EALs are included within existing IC RA2, and new ICs RS2 and RG2. Associated EAL notes, bases and developer notes are also provided.It is recommended that these EALs be implemented when the enhanced spent fuel poo1 level instrumentation is available for use.The regulatory process that licensees follow to make changes to their emergency plan, including non-scheme changes to EALs, is 10 CFR 50.54(q). In accordance with this regulation, licensees are responsible for evaluating a proposed change and determining whether or not it results in a reduction in the effectiveness of the plan. As a result of the licensee's determination, the licensee will either make the change or submit it to the NRC for prior review and approval in accordance with 10 CFR 50.90.3 2nEYv TERMunOLOnnGY nnaED In N-"n BO_^There are ceveral key term:n~ that appear throug~hout t.he NEL 99 01 methodology. These term":. a.e i-ntroduced': in hi:. ceetion toc upp.ort of :ube.guent material2. A: an aid to the reader, follow--ing table ic provide~d a: an ove'r.'i-,- to th'e relatincnhip c~f the Emergency Em...erge..cy Emergency Emergency'.Notes *--Nte --Notes '-Note:*---Basi 2.1 EMERCENCY CLAn~LrICATIoN LEVEL (ECU One of a Oct of name: or title: cotabliohed by the US Nuclear Regulatory Commi::ion (NRC) for grouping off normal event: or condition: according to (I) potential or aemal effect: or conoequonce:, and (2) reculting on:ite and offoite re:pon:e action:. The emergency ela::ification level:, in acending order of se~ ~rit~, ar~.U Notification of Unu:ual Event (NOUE)Ul-A-lei* itc. A~rea (^AE)£ General Emergency (GE)I -I ~TITIanhIvn flT I inUIIflI ,-.venr Ira fl I-. I lee o ... f safet" of the*. plant or indiate.... a......o.

ecurit threat to. facili.t" protection has been.*expected unle:: fu"rther degadation of :afct s~tm ocu" 4 Pu:rpesi:

T he purpoce ot tux: s classification is to assure that 1 t, to bring the operations staff to~e4~e0 4ii-fu~e a state of readiness, and to.~2I~-A4ei4 Events are in progress or have occu..rred -which involve an act+'ual or poten:tial s.botantial degradation cf the l....l of cafc, of +the -plant cr a c'eeu-' eve... tha ... pr...obatle plfe threatenin reede tort peretcnnen efrh puagc to HOSTeqiLmEn beAuCTO ofa rOuT iLE ACTON.Anyrelasc ar exectd t b liite toomal faetoncof he EPA PA...e .. ..... ...e .... .. .... .. .. ..... ... neplu e__;_Pupose: Te pup ose= + of' thic: claci..... o i c t....... asur ta eer-cypesnnl r 21. I General Emerzencv ,(GE), Evento are i:n progreoc or have occu"-red -which i.nvolve actual: or su.'bstantial ACTION that recul:to in an actual !boo of phycical control of the facili.ty. Releases can be reaoenabkl, expected to exceed EPA DPAGP expocure. leve+lsoffcit for more tha n the+the,.,ou li...... and offcite organizational measurement, to..initiate.additional.me..ures.as 5

2.2 INITIATIiNC

Co~inITLoN (IC)An event or condition that aligns with the definition of one of the four emergency clazoifloation levelo by virtue of the potential or actual effecto or consequences. Discussion: An IC describes an event or condition, the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed no a continuous, measurable parameter (e.g., RCS leakage), an ovent (e.g., an earthquake) or the status of one or more fission product barriers (e.g., boo of the RCS Appendix I cfNUREG 0651 doeo nct contain example Emergency Action Levelo (EAIo) for each ECL, but rather Initiating Conditions (i.e., plant conditiono that indicate that a radiological emergency, or evento that could lead to a radiological emergency, hao occurred). NUREG 0651 otateo that the Initiating Conditions form the baoio fc~eotablio~ent by a lieenoee of the specific plant ino~mentation readingo (go applicable) which, if cxceede~ would initiate the emergency clasoification. Thus, it io the specific inotrument readingo that would be the EALs.Considerations for the asoignment of a p~icular Initiating Condition to an emergency clasoification level are discusoed in Section 3.2.3 E~ncLNcY AC~ON LE~L (EAL)A pre determined, oite specific, oboer;ablc threshold for an Initiating Condition that, when met or exceeded, places the plant in a givcn emergency claosification level.Discussion: EAL otatemento may utilize a v~ety of criteria including inotrament readingo and otatuo indications; oboervable events; resulto of calculationo and analyocs;entry into particular procedureo; and the occurrence of natural phenomena.

2.1 FIsSIoN

PnoDucrr BARmEn TIIUESIIOLD A pre determined, oite opecific, obser.'able tI potential boo of a fission product b~cr.~rcofloIC meicating h los or Discussion: Fiooion product barrier threoholdo represent threats to the deI~nse in depth design concept that precludes the relenoc of radioactive fiooion producto to the environment. Thio concept relieo on multiple phyoieal barriero, anyone of which, it maintaincd intact, preoludeo the rcleaoe of oigniflcant amounto of radioactive fiooion produeto to the environment. The primary flooion product barriers arc: E-4**el-Gla4 U Reactor Coolant System (RCS)E-Ce~ai~en~ Upon determination that one or more fiosion product barrier threoholdo have been sxcs~de& the o~mhination of barrier boo and/or notential boo thrcoholdo is zomoared to the fiooion product b~ier IG~,AI criteria to detcrmi In some accident oequences, the ICo and EAt 6 ne thc appropriate ECL presented in the Abnc~mal Radiaticn L~eIci Radizlogizal Effiuznt (A) Rc~cgniti3n Categc~' ~'ill b~ zxcedad at th~ znm~ time, cr shortly after, th~ loz of on~ or mzrz fission product barri~r~. This r~dundanzy is intantftmal as thz fcrmer ICs addrzss mdioasti~'i~' rzIc~ses that rzsult in zzrtain zffsit: d~s~s fram whztz;zr zauoc, including events that might net be fully cnocmpassed by fissicn preduct barriers (e.g., spent fuel pool accidents, design centainment leakage fcllowing a LOCA, etc.).7 A-3 OFl lTrhi NEDl a990_1 EMEl~rGENCrY d~l CA~g25ItICTIONmlk SCrH~EME h 3.1 A:SICN1~U~NT or EMsncENv~' CLASSIrICATION LEVELS (ECLS)An effective emcraencv clacificaticn ceheme muct incorcorate a realictic and g1~iirnt~ n e~ment '~f rPdc h~th t'~ r.lnnt ~prkerc and th: hea~h-~~-sa~et~-rick: in Underectimating the pe er, there are alco rickc in overe: st-aiekg public. There are obviou: actual threat from an event cr the threat as well (e.g., harm condition; hcwcv: attempt: to ctrke an, a~pproprate betw;een reac.enably a:ntici:pated even~t ci There are a .-nge of "non emergency ev'ente" reported to the US N-o-le=" to determi..ne t.he attribute: of each ECL. The goal o~f thi: pro~s is tc an w'.er fc,,ll+'ow'ng

ource:

infc..-m.ation and c..nte.t f.,or the de..elopment. of ECL U Typical abnorm'=al and emergency operatng procedure etpoints and criteria U Typical Techn:ical Spcifcaton^ ,liio

Reniew ofelecedus sUpdaedt FialSer ' ........ s Report (F+o acidnt.nalc The follo'wing ECL attributes

-,e*re cerated by the Re-vicion 6 Pre:paration Team. to aid in include the attrbut:,.. in thi: ....io ince they' may be .......in.brie..n an training.. in thic docu'ment. fle attribute: or eaefl LLL ore precentec tetow.8 3.1.1 344- Notification of Unusual E vent O.TOUE) o-r unuIsua, t Lven, aS o.eiineo in secixon /. i. , o-.ur lo ,ni9-, :iimne.u L an event or condition tha involves: (A)A precursor to a more significant event Zr condition.(C) A consequence oth"r....e si agn~ificant enough t ......a..t notifcatio ... lal, tte an...d An Alert, as defined in section 2.1!.2, incluades bu-t is not limi~ted to an event or codiio ..a. inv...............l.......e.s:~tn a ia ,t control radiation levels w.ithin the plant, or a release of r.adioactive materia's to the 3.1.3 Site b=ea Emergency A Site Area Emergency, as defined in section 2.1!.3, incluades buat is not limited to ( A)"A laos or potential1 loss ofe+ any, product+ barriers c1ad, B)A peuro event....or conditi that, may. lead to the los or potentila loss of multiple , issio product barr..... withina.relatively..... period.of.tim ..Pecrsrvetsan conitsions --ofa this- ...pe include *°:"' those tha challeng themoitoin .......or .conto ot than.......f.an.EPA.P ...at or beyon..th site bounday 1 ACg'TION%. occu,,in within the.l- plant PRO+1D AEA 9

3.1.1 General

Emergency A General Emergency, ac defined in section 2. 1.1, includes bat is not limited to an, event or condition that in;'vels'e: pr.u.tba... r... Precur.or....ent..... conditions ofrtis- tpe include those that lead dirctly to c.. re. damage.. an loss of cortaimer.nt tha an. EP A ^P.G at or beyond the site boudar..4, (D)A HOSTILE ACTION rec.ulting in the o, ..f key ...... funtion..

(react..i..

-.,......., 3.1.5 Pick Info.rm.ed Insights i~mergency prepareJnesc xc a ~ietence in depth meacure tflat Ic nuependent 01 tile accessed rick from any particular accident sequence; ho;vever, the development of an effective emergency clascification scheme can benefit from a review of risk based acccccment recultc. To that end, the development and assignment cf certain ICc and EALs alco concidered incightc from several cite cpecific probabilictic cafety acccscmentc (PSA alco IGiown as probobilictic risk acceccment, PILA4 Seine generic incightc from thic review included: 1. A ~xd~nt '~guences involving a prolonged less of all AC power are significant ccntributorc to core damage frequency at many Preccurized Water Reactorc (PWRc)and Boiling Water Reactorc (BWRc). For this reason, a locc of all AC power for greater than 15 minutec, with the plant at or above Hot Shutdown, was assigned an ECL of Site Area Emergency. Precurcor eventc to a loss of all AC power were alco included as an Unucual Event and an Alert.A ctation blackout coping analyses performed in responce to 10 CFR § 50.63 and Regulatory Guide 1.155, Station B!aekoot, may be used to determine a time baced criterion to demarcate between a Site Area Emergency and a General Emergency. The time dimension is critical to a properly anticipatorj emergency declaration since the goal ic to minimize the time available for State and local officials to develop and implement offoite protective actions.2. For severe core damage events, uncertaintiec exist in phenomena important to accident pro~ecsionc leading to containment failure. Because of thece uncertaintiec, predicting the statas of containment inte~ity may be difficult und re ra ident conditianc. Thic ic why maintaining containment integrity alone following sequences leading to cevere core damage ic an insufficient basic for net escalating to a General~e~gei~ey~

3. PSAc indicated that leading contributorc tc latent fatalitiec were sequeneec involving a containment bypooc, a large Losc of Coolant Accident (LOCA) with early 10 e-.nt boozed ICs and Each type is discussed below.Symptom haz..ed ICs anrd E.A~sar"e or cond.itions are mea,.urable radiologieal effluent, etc.). When oneeor more of these param-eters or conditions, are off EAts that refer specifically to the level of challenge to the principal barriers against the rees of.... radioa.ti...mate

..al ..r.m.th.r.actor.core.to.the.environ ent.Te.ebrir ar- th"e'e+ fel ladnteractoa .r. .. coolan system a presr boudar, and.*... the.+ cont.inment Th...b....er.b...ed .....and .......consider ..the.le.el of ...hall ...e to each iniida br n" ...... 1 ....m~aoe nazar-as suen as a gas release.3.3 NSSS DESIGN DwrniE~clss The NE' 99 01 emergency. classificatio schem accounts... for the design differences betw;een PWP~s and BW?.-s by ....~,in EA.. s uni...ue to each... typ ci Dev'elopers --ill need to consider the relev'ant oopeets of their plant's design and site specific classification scheme. Th~e goal is to. maintai as much..... fi-lt as v.......The guidance in Nrct 99 01 is not applcable to. advanceed pasve light.. water...de vac'-eloped" i... aodac with.. NEIxr 0"7 01 Afthdo -g fr D.... op...... fE.'........ II 341.4 ORGANIZATION AND PRESENTATION OF GNIAiqEINFORMATION The scheme's-eei information is organized by Recognition Category in the following order.* R -Abnormal Radiation Levels / Radiological Effluent S4eedier. 5* C -Cold Shutdown / Refueling System Malfunction -Seectecn-7

E -Independent Spent Fuel Storage Installation (ISFSI) -4eletion8
F -Fission Product Barrier -Seetion-9
H -Hazards and Other Conditions Affecting Plant Safety-8eSeft-4n 1* S-M -System Malfunction S4ectionA4 The following inf..rmat...n..n.

guid..n...i. prvid..d........h..C. ECL the a:signecd emergency classification level fcr the IC.Operating Mded Applicabili-. Lit t ..he modes during which the IC and-If',- th ... g-en-ri' to the_. development of an example EPI cannot. b...e+ used (e.g., -an...umed instrumentation r...g. is net a.a.a. l the plat) the'* should .ttempt degee of bm'ier challenge (i.e., potentia l:ss ,or Thi.s presentation method shows..+the syn'ergism among tetrsodadspot cuaeassmns ofte ICan EA,^ s. In Tsom e ae, h assa includesea~ia~ ......... rlvan sour. in............ calculations, etc... D vepr n.... shold. nt...incudedin..e.ste.semerenc classification scheme basis Developers may elect to include inform,.ation £ ECL A~ssignmen~t Aftr'ib.tcs Lccate:d w,:ithin the Developer Notes scecti12 3J.1.5 IC AND EAL MODE APPLICABILITY The NEI 99 01 emergency classification scheme was developed recognizing that the applicability of ICe and EALs will vary with plant mcde. For example, seine symptom hazed ICe and EALs con be assessed only during the power operations, startup, or hot standby/shutdcwn modes of operation when all fission product barriers ore in place, and plant instrumentation and safety systems are fully operational. In the cold shutdown and refueling modes, different symptom booed ICe and EALs will come into play to reflect the opening of systems fcr routine maintenance, the unavailability of some safety system components and the use of alternate instrumentation. The following table shows which Recognition Categories are applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes.MODE APPLICABILITY MATRIX Category Mode R C E F H SM Power Operations X X X X X Startup X X X X X Hot Standby X X X X X Hot Shutdown X X X X X Cold Shutdown X X X X Refueling X X X X __Defueled X X X X 13 Operating Modes Technical TABLE 1.2-. --COmmented [DWS1]: VI TS Table 1.2 Mode Defleition MODE Reactivity % Rated Thermal Average Coolant Condition. Keff Power* Temperature

1. Power Operation 2. Startup 3. Hot Standby 4. Hot Shutdown 5. Cold Shutdown 6. Refuleling**

NA Defueled> 0.99> 0.99< 0.99< 0.99< 0.99 NA> 50%< 5%0 0 0 0_> 350 0 F> 350°F_> 350°F 350 0 F > Tavg >200 0 F< 200 0 F< 140 °F All fuel removed from thle reactor vessel (full core offload during refueling or extended outage)*Excluding decay heat.**Fuel in the reactor vessel with the vessel head closure bolts less than fully' tensioned or wvith the head removed.Pe r Operat en (1) Reacte ...... r 5.... Koff ¢-Strup (2):*, ...a. to, ",. wer 5t%, Ke-- 1.6StadbB(3:ISS 35OCUMENT- .9 proidnginfrmngbakgrun adRevuelipmngt6) Onento in headil acelsuib forma. Itcan e reerre to n bri ltn lsiutos tand whlyen maionednatleerec cle asisiiainifecsa.Th document is anlnerasat ofa emerencu larssabifhic gcniuationsce .Th mnanagemaent controls for EP-related equipment and explaining an emergency classification to offsite authorities. The content of the basis document includes:*A site-specific Mode Applicability Matrix and description of operating modes (see Section 1.5).14

A discussion of the emergency classification and declaration process (see Section 2).* Each Initiating Condition along with the associated EALs or fission product barrier thresholds, Operating Mode Applicability.

Notes and Basis information (see Sections 3-8).* A listing of acronyms and defined terns (see Appendices A and B. respectively). A basis section should not contain information that could modify the meaning or intent of the associated IC or EAL. Such information should be incorporated within the IC or EAL statements, or as an EAL Note. Information in the Basis should only clarify and inform decision-making for an emergency classification. Basis information should be readily available to he referenced, if necessary. by the Short Term Emergency Director/Site Emergency Director (STED/SED). For example, a copy of the basis document could be maintained in the appropriate emergency response facilities. Because the information in a basis document can affect emergency classification decision-making (e.g.. the STED/SED refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q).I1.7 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA The criteria/values used in several EALs and fission product barrier thresholds may be drawn from AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments. Appropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to deternine if an evaluation pursuant to 10 CFR 50.54(q) is required.15 I SITE-SPECIFIC SCHEME DEVELOPMENT GUIDANCE This ecti. prvdsCeale uncep~z!,Leapoc fr d....l.ping ... sit .p.. ii .. m...rg...ny .l...iicat.. ....r g ..n ..y -pr t n p r o. .d.r. g en.. .r i. m at. r ia l p r e p a d b y ......... .,en d o o w ;n e r g r o u p s is c o n v e rte d b y e a c h ...u l r p o.. .r.. l. n t i n to: ...: ...t .sp e ifi ..m .rg e ... *pr t n r e d r Likewise, the emergency classificatio ..h.. develop.. r I .. ll. us1 ..h e generic guidanc ... NE TIT It is important that the NEI 99 01 emergency classification scheme be i'mplemented as an integrated package.. Slcted u.. se.. of portions of this guidanceo is strongly, disc...ur.ged4 as it ....l l.ad to. an in.n...en or4- incompl..

t. eme...rgency classificatio

..h. m tha will. l.;11ikely net r...eiv.4.1 CENEnAL IMPLEMENTATION CUIDANCE The g..uidance. in NE! 99 01 is n..t intendedn to b. .pplie tn plats+ ,as is";.generic guidance as p.sib.. Th li ometteitn ..f. th generic..n.t.a..ng chaPracteIstics. l.cale, plan desima, operati-ng features, te....i.-noloy et. Meetig thi ge lwll rslt in a sherter ,andlaa4 h les ubro rac, NRC rev++h.iew and approval ..... e,4 a,: o,,sessahle. As discu...sed in Section 3, the generi guidance:, incl:udes,4o Ic and E ^' °se.'ves a. spec7ific prpose...The....i..the.f.nd...ntal.e.ent er condition.. requi..ng c If++ some. featu+...re" of.. .........o .. or d ..sign. is+ ne comptibl with a generic IC er If-,an IC or EPJL incluades explicit reference to a mode dependent tec!=ieal spciiation that is not applicable to the plan÷ then- t+hat ICm an-,4In E2AI nee .net be hnclued inhe sitespecific schnteme. InThese worsc -dtevelpes mustproidentfe ad equatset clsification, ..heme (i re. LLCPr S).nTed 4.::........ in Appendix AAcemsadbreitiosSt 16 precentaticn pu:rpoece (e.g., "with a General Emergency at the* left'top of a user aid, follewed by Site Area Em..rg..n.y, .l.rt an NOUE).* M may be 'used in lieu- ef S* The IC: and EAL:s from Recognition Categoriec S .anC ma be. inorpor... ted int :..common precentation method (e.g., ....e .ale provided.that.al..related.n.t..... £ The. and EAL, for Emergency Direetor judgument an .....it relat.ed e..ent. may£ The term:n EAL and th~rechold may be used interchangeably. clascificatio scheme basics document.4..... TtcAL Cnni. .canisT'c charactericstics arc !ited below.£ The IC:, EAL:, Operatin Applica':'bilt citeri,% Note: n Ba- i information..:^ U hEA IC:, EALnt Operajeting"oe Arpplicabilid criter','~ Notles.n accifomt!17 i;;'hnere nem tc iltts prog n owgaig ttl"y.gnc lsitcto U z. The facilitates classification of multpe ...n.u...nt....nt. or condfitins, 1.3 INSTflUMENTAflON UCED ron EAL~,n tr... ntation ref.r. n e in.. EAI: s^tat..m.nt. should include described in the Technical, pe.cification or DMIRTScontr,.l requirement, nor po'wered from an Scheme. devl.. r , should ensur.... tato .p..fi. value"- .. used as EA ^s etpcints are -"ithin t.he calibrated range of the referenced ins~amcn~tatn, and cons~ider any automatic and v'iolations related to, EAL instrmnentation issues may be located on the NPRC -website.1.1 PREaENTATION or SChEME INrOIIMATtON TO U~Ens within: 15 minutes' afte the aviablt of indication~s tocln prtr hta pre.ent.tio method. s bet. ..upp..... theen users by facilitating accurate and timely.......... classification Tc this end, developers should conide the follow-ing points....Th..fi ..t.use ...of.an emegec cl.. si...cation. procedure are the oeramtor, in the.respnsibli. to,:. perf.rm other. critical tasks , and. wil likeyhav minimal..... assstan:eo in makldng a classification assessment.

,^ an emergency situation ev-les-e, membes of the Control Rooem staff are likely to* Emergency Directors in the TSC and/or EOF -will have more opportunity te' foe'-u on m.. ing anemrenycl.ifctin and........................probably...ha.e a visrs frmOer.atin avaiableto elp hem 18 Diretctrs and.'or Offsite R~esponse Organization personnel.

A "A,vubu°.d is an a:em~ presenmtatn me.ke pro;'.'ded +" tcz*n all+e..nclu..:..h...C.. Operat.ng Mo..... ppli..ab.l.ty .......a, ......d. ot....N.. t... may be wallboard bu-t it shoul:d be readily to emergentcy clas.sification decisica maker:.AtRnaie prsn Cation. method: fcr the Rec + gnition Cat... g..... F I:, .ad fis.ion... prouc to promote consistenc~ across mc Inoustry.1.5 I!mcRATION' or IC°JEALg ".rrTn PLANT PRC"CED'JRES A rigorous integration of IC and EAL references into plant operating procedures is not recommended. This approach would greatly increase the adminis~ive controls and workload for maintaining procedures. On the other hand, pefformance challenges may occur if recognition of meeting an IC sr EAI is based solely on the memory of a licensed operator or an Emergency Director, especially during periods of high stress.Developers should consider placing appropriate visual cues (e.g., a step, note, caution, etc.) in plant procedures alerting the reader/user to consult ~e site emergency classificatisn procedure. Visual cues could be placed in emergency operating procedures, abnormal operating procedures, alarm response procedures, and normal operating procedures that apply to cold shutdown and refueling modes. As an example, a step, note or caufion could be placed at the beginning of an RCS leak abnormal operating procedure that remind: the reader that an emergency classification assessment should be eifei~ied 7 4.6 BAsIS DOCUMENT A bais* doumnti an inera arfaneerec elassitication scheme. the matcr:al in tnin aecument supports proper emergency ciassuncanoen aeesion ma~rng ny provucin:g can orefcre to in' training situation and. whn m., king. an" actu'al emergency 19 a minimum, the following:

A site cpecific Mode App'icability Max a znd decoription of operatingcimilar to mat precentee in ceenon iz.£ A diccuccion of the emergency elaccification and declaration preceec reflecting the material precented in Section 5. Thic material may be edited ac needed tc align with cite epecifie emergency plan and implementing procedure reguirementc.

£ Each Initiating Condition along with the accociated EALe or ficcion product barricr thrcchcldc, Operating Mode Applicability, Notoc and Bacic informatien. £ A licting of acrenyme and defined termc, cimilar to that presented in Append~cec A and B, respectively. Thic material may be edited as necded to align with cite speeifie eha~eteristie~

~ specific background or tcchnical appendices that thc devLp~r.~bJi~...

weuld be useful in explaining or using element of thc emergcncy classification sehe~e A Basic section should not contain information that could modify the meaning or intent of the acceciated IC or EAt.. Such information chould bc incorporated within thc IC ot LAd. statements, or as an EAL Note. Infcrmation in the Basic should only clarify and inform decision making fcr an emergency classification. Bacis informatien should be readily available to bc referenced, if necessary, by the Emergency Dfreetor. For exumple, a copy of the basis document could be malntair.ed in the appropriate emergency response facilities. Because the information in a basis document can affect emergency elacsifieation decision making (e.g., the Emergency Director refers to it during an event), the NRC staff expect that changes to the basic document will be evaluated in accordance with the provicionc ef 10 CFR 50.51(q).1.7 EAL'TtrnESIIOLD RErERENCLIS TO AOP AND EOP SETPOINT~ICIUTEJ1IA Ac reflected in the generic guidance, the criteriaivalues used in several EALs and ficcion product barrier thresholds may be drawn from a plant's AOPc and EOPc. This approach Ic intended to maintain good alignment between operaticnal diagnoces and emergency elacoification assescmentc. Developere sheuld verify that appropriate administrative controls are in place to encure that a subsequent change to an AOP or EOP ic cereened to determine if an evaluation purcuant to 10 CFR 50.51(q) ic required.1.8 DEVELOPEB AND UTEEn FEEDBACK or comments concernring tne material in........... ay....... to t tn NEL merenc Prpardnec s, NIU .......force ......r. or cubmired~g to the ETmer c--ar Pea dns rq nty Acked Que......nc ....... ........ ..... ...20 62 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 54-2.1I GENERAL CONSIDERA&TIONS When making an emergency classification, the Emergencey DirectorSTED/SED must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL.NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, Interim Staff Guidance, Emergency Planning for Nuclear Power Plants.All emergency classification assessments should be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The validation of indications should be completed in a manner that supports timely emergency declaration. For ICs and EALs that have a stipulated time duration (e.g.,* 15 mi.n...,,,: 30.. minutes,,, t..)...the Emergency DirecterSFEI)/SEI) should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that l) the activity proceeds as planned and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements ofl10 § CFR 50.72.The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dece ........... chemict.r' zampling, RCS !cak rate calculatien, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC 21 expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., the. ........ , ' en ehift).While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 scheme provides the Emergneny DireeturSTED/SEI) with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emer~gencey will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.&2.2 CLASSIFICATION METHODOLOGY To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with plant procedures. When assessing an EAL that specifies a time duration for the off-normal condition, the"clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01.$---32.3 CLASSIFICATION OF MULTIPLE EwENTs AND CONDITIONS When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared.For example:* If an Alert EAL and a Site Area Emergency EAL are met, whether at e~ne u.nit e~r at..t, ..different uni,,,* a Site Area Emergency should be declared.There is no "additive" effect from multiple EALs meeting the same ECL. For example:* If two Alert EALs are met, whether at..... uni er at... twe .iffer.n .uni an Alert should be declared.Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events.CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is 22 declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original exent or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. [or events that occur in Cold Shutdown or Refueling, escalation is via EAI~s that are applicable in the Cold Shutdown or Reftteling modes, even if H~ot Shutdown (or a higher mode) is entered during the subsequent plant response. Inl particular. the fission product harrier EALs are applicable only to events that initiate in the Ilot Shutdown mode or higher.Onemadiffaernt mo dei ece. n e events or condition, nhat ruldated to thein oriecedin n AL withi a renditiven, shrtquiring oemergien., lasfchatgein hudbe ECaLuatsdIMgINENT)th If, i and jugent aolialft the Eperaeing mrcderatthE/SD tmeftetinew avntEA ir IMIENTiton apliabergincy clasfctonsold beto~o madueln mass c if heELHasote meut.o (hir appligabe mode ll enterged curingfthatiubseeuentplant rppponaeh In particularly thpofisntitntherhduct barrergency carssaipicabinlevenly tsivnte tht poineditiatnai thme Hot Shutpwnmdentto ofCLMsERGAENC OLSSFICAMIONLENT L C UPGRADI~NGs ONRDN reAin EC al er townrddweh events or conditions thatcolledt meetinso excedhinghet andEA withino arlatively eshots pnterio ofsime-(ipe.,achangeading theqECLirmns I remeNT. Ifi tejdowgmedntgo the EmLi erency apprcopriateD/,th meetingL anoELdi IET then ebsdo emwergec cassificatleionshould bEAs) made ECasyfh EAasosmy bee mt.eWhileappliabl Toe aolloemrgncycasfctonlvlti approach tdonrigoremisn articularELy importammntdttedhihe emergency CLasfcto leessic ctpoidnWes aditonalitime for imLemntatio ofst protectiveomeasures d lrowngradigteELidemdaporteo thermnew teCwol then bmer bae d on a acracwihpatprocedures. Site Area Emergency with no Downgrade or terminate the emergency in long-term plant damage accordance with plant procedures. 23 Site Area Emergency with Terminate the emergency and enter recovery in long-term plant damage accordance with plant procedures. General Emergency Terminate the emergency and enter recovery in accordance with plant procedures. is pravided in P.!S 20037 02.&--2.7 CLASSIFICATION OF SHORT-LIVED EVENTS^.As di..u... in S.ectn:.2, a c Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed e.... Ifa

n o...ur. thatmet cr ........ an E. A , .......t.......mu.t.b.....lar

.radls ..f.. it.. continued presence at the time of deelo,,'a,:iein.Examples of such events include a failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip or an earthquake.CLASSIFICATION OF TRANiSIENT CONDITIONS Many of the ICs and/or EALs contained in this document employ time-based criteria.These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a.. fe ........ a fc,,... ... The following guidance should be applied to the classification of these conditions. EAL momentarily met during expected plant response -In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. EAL momentarily met but the condition is corrected prior to an emergency declaration -If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example.An ATWS occurs and the auxiliary feedwater system fails to automatically start.Steam generator levels rapidly decrease and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal condition prior to an emergency declaration, then the classification should be based on the ATWS only.It is important to stress that the 15-minute emergency classification assessment period is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event;emergency classification assessments must be deliberate and timely, with no undue 24 delays. The provision discussed above addresses only those rapidly evolving situations where an operator is able to take a successful corrective action prior to the Emiergency Di-ree-tfSTED/SED completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1 022 is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. Gui:danee on the rec-raeticn uf an. emergency deelaratien rep~ed th e NRC is dss-ezsed 25 63 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS GENERAL SITE AREA EMRENY EMRENYALERT UNUSUAL EVENT RG1 Release of RS1 Release of RA1 Release of RU1 Release of gaseous radioactivity gaseous radioactivity gaseous or liquid gaseous or liquid resulting in offsite resulting in offsite radioactivity resulting radioactivity greater dose greater than 1,000 dose greater than 100 in offsite dose greater than 2 times the (-site-mrem TEDE or 5,000 mrem TEDE or 500 than 10 mrem TEDE ...: *...mrem thyroid CDE. mrem thyroid CDE. or 50 mrem thyroid Op. Modes: All Op. Modes: All CDE., .... *Op. Modes: All limits for 60 minutes or longer.Op. Modes: All RG2 Spent fuel pool RS2 Spent fuel pool RA2 Significant RU2 UNPLANNED level cannot be level at (site-speecihl lowering of water level loss of water level restored to at least -,.esei -.patien;"l .5 above, or damage to, above irradiated fuel.(,,t" ......i Lwc";:'z 3., .. ft. (Level 3)1. irradiated fuel. Op. Modes:" Allft., Op. Modes:" All Op. Modes:" All (Level 3)1 for 60 minutes or longer.Op. Modes:" All RA3 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, shutdown or cooldown.____________________________Op. Modes: All-Commented [DWS3]: V2 SFP Levels Drawing-Commented [DWS2]: V2 SFP Lneves Drawing 26 RG1 ECL: General Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE.Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)Notes:* The STED/SEDEmerga:noy Di-ectzr should declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.* The pre-calculated effluent monitor values presented in EAL # 1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. (1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Monitor Reading(WRGM rate) 2.85E+8 Time After Shutdown Reading______________ <ihr If>1hrto <2hrs RM-6481-1l* (M SL A ) 1310 mR/hr 1060 tnRihr RM-6482-1l* (MSL B) 1310 mR/hr 1060 mR/hr(MSL C) 1310 mR/hr RM-6481-2' (MSL D) 1310 mR/hr j 1060 mRlhr*With release path to the environment firom affected steam line, open ASI)V or SRV, line is faulted. or open steam supply to 1 -F W-P-37A. (site~w..; .....fi -mzn-it.cr lizt- anda .. ..... " c°)OR (2) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond (sitc apezifiz dcc: receptcr pzint~rhe site boundary]. OR (3) Field survey results indicate EITHER of the following at or beyond (site zpe.ific~d~r......... .pointte site boundary]: C ommentedl [DWS4]: V3 EPCALC-06-02 -Effluent Monitor Values for R EAts S- -] Commented [DWSS]: V3 EPCALC-06-02 -Efun oio' Values for R EALs 1 Commented [DWS6]: v4 0DCM and TS Basis for Site/Boundary Receptor Point S- Commented [DWS7]: V4 0DCM and TS Basis for Site[Boundary Receptor Point Closed window dose rates greater than 1000 mR/hr expected to continue for 60 minutes or longer.Analyses of field survey samples indicate thyroid CDE greater than 5000 mrem for one hour of inhalation. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both 27 monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classify.ing events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.28 RG2 ECL: General Emergency Initiating Condition: Spent fuel pooi level cannot be restored to at least-: ..s.t .epecific*' Leve 3.. , 1.5 ft. de....ef... jden-Lvel

3) for 60 minutes or longer.Operating Mode Applicability:

All Emergency Action Levels: INote: The Emergency DireetorSTED/SED should declare the General Emergency promptly I upon determining that 60 minutes has been exceeded, or will likely be exceeded.I (1) Spent fuel pool level cannot be restored to at least (site epecific Level 3 ;*alue~I.5 ft.above the fuel racks- for 60 minutes or longer as indicated by SF-LI-2616 (MPCS computer point A4172) or SF-LI-2617 (MPCS computer point -{ Commented [DWS8]:-V2SFP Levels Drawing Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment. lPost-Fukushima order EA-P1-051 required the installation of reliable SEP level indication capable of identifying normal level (Level I), SEP level 12 ft. 3 in. aboxe the top of the fuel racks (Level 2) and SEP level 1.5 ft. above the top of the fuel racks (Level 3).The Spent Fuel Pool Instr-umentation System (SFPIS) consists of two newv independent level instrument channels (SF-L-2616 and SF-L-26 17) in the Spent Fuel systcm. The SEPIS channels will be used to monitor spent fueI pool level during and Ibllowing beyond design basis events that could challenge the capability to ensure optimum protection for the stored fuel assemblies in the pool.Each channel is capable of measuring SEP level over a span from just above the top of the spent fuel racks to the normal SEP operating water le~el. The SFPIS will be monitored in accordanceBeyond Design Basis guidelines contained in FS~s for Extended Loss of AC' Power and Alternate SFP~ Makeup and Cooling.It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity. 29 RS1 ECL: Site Area Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE.Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)Notes:* The Emzx~rgoncy DirectorSTED/SED should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. (1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Reading RM-6528-4 (WRGM rate) 2.85E+7 uCilsec Time After Shutdown Reading<lhr >lhrto <2hrs RM-6481-1' (MSL A) 130 mR/hr 100 mRlhr RM-6482-1' (MSL B) 130 mRihr 100 mR/hr RM-6482-2' (MSL C) 130 mtlhr 100 mRihr RM-6481-2' (MSL D) 130 mR/hr 100

With release path to the environment from affected steam line, open ASDV or SRV.line is faulted, or open steam supply to I-F W-P-37A.OR (zitc zpecific monitor list and: threzho!d v'alues)(2) Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond (zitz specific ...... r ....p..... p-oinv.,the:'

site'boundary4 OR (3) Field survey results indicate EITHER of the following at or beyond (4ite-specifie-dese ... v-. v... the site boundary.Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer."- Commented EDWS9]: V3 EPCALC-06-02 -Effluent Monitor/ Values for R EALs I Commented [DWSIO]: V40ODCM and TS Basis for Site[Boundary Receptor Point [DWS11]" V40ODCM and T5 Basin for Site Boundary Receptor Point Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation. 30 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via IC RG I.31 RS2 ECL: Site Area Emergency Initiating Condition: Spent fuel pool level ... ite ,,".. 'pci Lee 3 d'aezeriptic'nl.5 ft.kLevel

3) ---Commented

[DwS12]:VSF v eisr r L wviO s Operating Mode Applicability: All Emergency Action Levels: (1) Lowering of spent fuel pool level to (site: ......i ... 3 ft above the fuel racks as indicated by SF-L1-2616 (MPCS computer point A4172) or SF-LI-2617 (MPCS computer point A4220).-). -. Commented rDWSX3]: V2 SFIPLeves Draing Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IiMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Post-Fukushima order EA-12-051 required the installation of reliable SEP level indication capable of identifying normal level (Level 1). SEP level 12 ft. 3 in. above the top of the fuel racks (Level 2) and SEP level 1.5 ft. above the top of the fuel racks (Level 3).The Spent Fuel Pool Instrumentation System (SFP1S) consists of two new independent level instrument channels (SF-L-2616 and SF-I_-2617) in the Spent Fuel system. The SFPIS channels will be used to monitor spent fuel pool level during and following beyond design basis events that could challenge the capability to ensure optimum protection for the stored fuel assemblies in the pool.Each channel is capable of measuring SEP level over a span from just above the top of the spent fuel racks to the normal SEP operating water level. The SFPIS be monitored in accordance wvith Beyond Design Basis guidelines contained in FS~s for Extended Loss of AC Power and Alternate SEP Makeup and Cooling.It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity. Escalation of the emergency classification level would be via IC RG1 or RG2.32 RA1 ECL: Alert Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4)Notes:...The Emerg.n a .....torSTED/SLI) should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. (1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Monitor Reading RM-6528-4 (WRGM rate) 2.85E+6 uCi/sec RM-6481-1" (MSL A) It) mR/hr RM-6482-1

(MSL 13) 10 mRlhr RM-6482-2" (MSL+ C) 10 mR/hr IRM-6481-2' (MSL D) It) mnRih*With release path to the environment from affe~cted steam line, open ASDV or SRV. line is faulted, or open steam supply to 1-FW-P-37A. (C:tz e-ecific mzn:^-ite

....tx dthrce^cld OR (2) Dose assessment using actual meteorology indicates doses greater than 1 0 morem TEDE or 50 mrem thyroid CDE at or beyond (site. epeci'c dccc rceptor point)~he site boundar,,j. OR (3) Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site-s~eeifie ...... e site bounday fr one hour of exposure.OR (4) Field survey results indicate EITHER of the following at or beyond 5ee recetor e~nthe site bo~undar)]:

Closed window dose rates greater than 10 mRlhr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.

I Comment~ed [DWS14]: V3 EPCALC-06-02 -affluent Monitor Values for R EALs SC ommntedl [DWS15]: V4 0DCM and TS Basis for Site Boundary Receptor PointCommented [DWS16]: Vs40ODCM and TS Basis for SitelBoundary Receptor Point l Commented [DWS17]: V4 00CM adT5BSi for Site SBoundary Receptor Point 33 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a zignificant uncc~ntrclled Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via IC RS 1.34 RA2 ECL: Alert Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel.Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)(1) Uncovery of irradiated fuel in the REFUELING PATHWAY.OR (2) Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by high-alarm, or reading in excess of the current high-alarm setpoint on ANY of the following radiation monitors: I /M-6518-1 FSB High Range RM-6562- I FSB Vent IRM-6535A-1 Manipulator Crane RM-6535B-l1 Manipulator (3) Lowering of spent fuel pool level to (cite specific L... 'e 2 ft. 3 in. above thc fuel racks on SF-LI-261 6 (MPCS computer point A41 72) or SF-Ll-261 7 (MPCS computer point A422tI). [See n ... t ... r ....sl Basis: PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal.[This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool (see-De'el~per Noes-). These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-l44--EL' 1.Escalation of the emergency would be based on either Recognition Category R or C ICs.EAL # 1 This EAL escalates from RU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g.,......t fr .......n.. r ...m.... imagec), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used-(e~g~., abei.!-off eiui'ie). Classification of an event using this EAL should be based on the totality of available indications, reports and observations. While an area radiation monitor could detect an increase in a dose rate due to a lowering of water 35"- Commented [DWS18]: V5 UFSAR Table 12.3-14 -CTMT tPost-LOCA Range-( Commented [DWS19]: V2 SFP Levels DrawingComamenteda [DWS20]: V6 Refueling pathway RU2 RA2 level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.EAL #2 This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event e.g., a fu..'l h....lin. ac, i ......., EAL #3 Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.Post-Fukushima order EA-12-051 required the installation of reliable SEP level indication capable of' identifying normal level (Level 1), SEP level 12 ft. 3 in. above the top of the fuel racks (Level 2) and SFP level 1.5 ft. above the top of the fuel racks (Level 3).The Spent Fuel Pool hnstrumentation System (SFP1S) consists of two new independent level instrument channels (SF-L-2616 and SF-L-2617) in the Spent Fuel system. The SFPIS channels w~ill be used to monitor spent fuel pool level during and folloxx ing beyond design basis events that could challenge the capability, to ensure optimum protection fbr the stored fuel assemblies in the pool.Each channel is capable of measuring SEP level over a span from just above the top of the spent fuel racks to the normal SEP operating xvater level. The SFPIS be monitored in accordance with Beyond Design B~asis guidelines contained in FSGs for Extended Loss of AC Po\ver and Alternate SEP Makeup and Cooling.Escalation of the emergency classification level would be via ICs RS1 or RS2-{aee--/?=2 36 RA3 ECL: Alert Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, shutdown or cooltdown. Operating Mode Applicability: All Emergency Action Levels: (1 or 2)INote: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. (1) Dose rate greater than 15 mRlhr in ANY of the following areas: Control Room RM6550 Central Alarm Station (CAS) by survey Secondary Alarm Station (SAS) by survey fothef OR (2) An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any of the following plant rooms or areas: related m-^de .... lizabili identified)Table H] 1)Area Mode Primary Aux Building 25 ft elevation 1 .3 7 ft elevation-26 ft elevation Turbine Building 1. 2. 3 Switchgear Rooms Essential 1, 2,3, 4 Non-essential Steam and Feedwater Pipe chases 1, 2. 3 Waste Process Building 25 ft elevation1,23 -3 ft elevation1,23 -31 ft elevation Containment

3. 4 Equipment Vaults 3. ,4

[DWS21]: V70~ps procedures 37 Basis: UNPLANNED: A parameter 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.IMPEDE; Entry into an area requires extraordinary measures to thcilitate entry of personnel into the affected room/area by installing temporary' shielding, requiring use of non-routine protective equipment. or requesting an extension in dose limits beyond normal administrative limits.This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The ..m. erg....nc... Dir......STEDiSED" should consider the cause of the increased radiation levels and determine if another IC may be applicable. For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g,, ..tal.ing tem...orar ..h......ng,reuig

u. e ... f non... r.. utin... pr..t..ti..

e-quipm...nt, requectin~g an....n....n..............d nrma dm .....trativ ........t).An emergency declaration is not warranted if any of the following conditions apply.* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.* The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility ofaromo are (.g.,. radi..grah, ofpanroomltrrarr rez.n., ...f.r,-et,,).

The action for which room/area entry is required is of an administrative or record keeping nature (e.g., no rm',al roundes or routine inepection).
The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.38 RU1 ECL: Notification of Unusual Event Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the (ie specific ....u.nt............tr....ng de......... ODCM limits for 60 minutes or longer.Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)Notes:* The .m.rg.n. Di... e......STED/SED should declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.(1) Reading on ANY non-isolated effluent path radiation monitor greater than 2 times the (cte .p..... effluent ...... centro.... g u' ccu .......)ODC M limits for 60 minutes or longer:(WTT Disch)RM-6521I-1 (TIB Sump)RM-6519-1 (SC B lowd own)RM-6473-l (WT LIQ kRM-6528-4 (WRGM [DWS22]: V32 UFSAR Table 11.5-1 [DWS23]: VS 012_Table 03 UFSAR WRGM/ Ranges OR (bite meniter lict nr.d thre~hcUd

.-lucez rcrr~ezndinez te 2 tim.ec the zcntrotlinr (2) Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.OR (3) Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the (cite epecific efflunt ....... e..n t.r........

d.. oc--.cnt)O...... V limits for 60 minutes or longer.Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (.g.,.an........le relace)... It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.39 Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.Radiological effluent EALs are also included to provide a basis for classify'ing events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.EAL # 1 -This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.EAL #2 -This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL i4-;'l associated with planned batch releases from non-continuous release pathways EAL #3 -This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways-feg. -spi4ls-ef ............. d ino zor ,dr........,.......... ev'whanger l!elkage in river "wter eystem,z et.)....Escalation of the emergency classification level would be via IC RA I.40 RU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel.Operating Mode Applicability: All Emergency Action Levels: (1) a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following: AND b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors:, j [DWS24]: V9 SFP Level fM-6535-A-1, Containment Manipulator Crane RM-6535-B-1, Containment Manipulator Crane RM-6549-1, FSB Spent Fuel Range Lowv RM-65 18-1, FSB Spent Fuel Range HiIcommented [DWS2S]: Vio UFSAR Table 12.3-14 -CTMT I Post-LOCA Range (sue Basis: UNPLANNED: A parameter change or an event that is not l) the result of an intended e~olution or 2) an expected plant response to a transient. T'he cause of the parameter change or event may be knoxxn or unknown.REFUELING PATHWAY: The reactor refueling cavity', spent tfuel pool and fuel transfer canal.This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnelfrom a rcfueling crew;) or video camera observations (if available). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations. The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level.A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.Escalation of the emergency classification level would be via IC RA2.41 74 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS Recognition Category "C" Initiating Condition Matrix GENERAL SITE AREA EMRENY EMRENYALERT UNUSUAL EVENT CG1 Loss of reactor CS1 Loss of reactor CA1 Loss of reactor CU1 UNPLANNED vessel/RCS inventory vessel/RCS inventory vesselIRCS inventory, loss of reactor affecting fuel clad affecting core decay Op. Modes: 5, 6 vessel/RCS inventory integrity with heat removal for 15 minutes or containment capability, longer.challenged. Op. Modes: 5, 6 op. Modes: 5, 6 Op. Modes: 5, 6 CA2 Loss of all CU2 Loss of all but offsite and all onsite one AC power source AC power to to emergency buses for emergency buses for 15 minutes or longer.15 minutes or longer. Op. Modes: 5, 6, Op. Modes: 5, 6, Defueled De fueled CA3 Inability to CU3 UNPLANNED maintain the plant in increase in RCS cold shutdown, temperature. Op. Modes: 5, 6 Op. Modes: 5, 6 CU4 Loss of Vital DC power for 15 minutes or longer.Op. Modes: 5, 6 CU5 Loss of all onsite or offsite communications capabilities. Op. Modes: 5, 6, De fueled CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.______________ ______________Op. Modes: 5, 6 42 CG1 ECL: General Emergency Initiating Condition: Loss of reactor vessel/RCS inventory affecting fuel clad integrity with containment challenged. Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)SNote: The Emergency DirectcrSTED/SED should declare the General Emergency promptly I upon determining that 30 minutes has been exceeded, or will likely be exceeded.I (1) a. RVLIS Full Range < (-141.5 in) PReacter ;e'eseL/RC$ level lezs then (site....eitie -4.... ) for 30 minutes or longer.AND b. ANY indication from the Containment Challenge Table C2. (see-hele..) OR (2) a. Reactor vessel/RCS level cannot he monitored for 30 minutes or longer.AND b. Core uncovery is indicated by ANY of the following: RM-6535A-1 (Manipulator Crane) reading greater than 9500 mR/hr IRM-6535B-t (Manipulator Crane) reading greater than 9500 mR/hr(-ite-Erratic source range monitor indication UNPLANNED increase in Sumps A orPstesefi ...ee4efank... -levels of sufficient magnitude to indicate core uncovery.(the .it..... ci ............. nVisual observation. AND c. ANY indication from the Containment Challenge Table C2. ......... ,.Containment Challenge Table C2 CONTAINMENT INTEGRITY not established

Kontainment H 2 concentration

_ 6°i4(Explcsi-.e mixture) exists inside centainment UNPLANNED increase in containment pressure* If CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.Basis: UNPLANNED: A parameter 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 unknowvn.-{Commented [DWS26]: Vii EPCALC-06-04 -RVLIS Values I Commented [DWS27]: Vt0 UFSAR Table 12.3-14 Manipulator Crane Monitor Range I" Commented [DWS28]: VI3 Containment Sumnps d[ Commented [DWS29]: VI4 1-12 concentration in containment 43 CONTAINMENT INTEGRITY: The procedurally defined conditions or actions laken to secure containment and its associated structures. systems. and components as a functional barrier to fission product release under shutdown conditions. This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.With CONTAJIvMENT INTEGRITY not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen bumn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment, If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. Manipulator Crane setpoint of 9500 mR/hr is 95% of the monitor range.In EAL, 2.b, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncoveiy has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor reactor vesseliRCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS. RVLIS LEVEL VESSEL LEVEL (%) (inches from vessel____________ flange)-108 119.8 100 81.3 90 i 31.8 80 j -17.7 44 70 -67.2 63 -101.9 RC-LI-9405, RC-LIT-9467, 60 -116.7 and the Tygon Tube do 55 -141.5 not indicate reactor vessel 50 -166.2 level when actual level is 40 -215.7 less than -95" due to the 30 -265.2 weir on the RCP 20 -314.7 discharge. 10 -364.2 0 -413.7 _________These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal;SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industfry Actions to Assess Shutdown Management.

Reference:

FSAR Table 12.3-14 45 CS1 ECL: Site Area Emergency Initiating Condition: Loss of reactor vessel/RCS inventory affecting core decay heat removal capability. Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2 or 3)Note: The Emergency DirzztzrSTED/SED should declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.(1) a. CONTAINMENT INTEGRITY not established. AND b. Reatere eLW C lve !es &"q RVLIS Full Range < 63% (- 101.9 in) 4 commented [DWS30]: VI1 EPCALC-06-04 -RVL1S Values OR (2) a. CONTAINMENT INTEGRITY established. AND b. IRVLIS Full Range < 55°/4 (-141.5{Commented [DWS31J: VII EPCALC-06-04 -RVLIS Values in u-.eactcr izevel iess tan rsite OR (3) a. Reactor vessel/RCS level cannot be monitored for 30 minutes or longer.AND b. Core uncovery is indicated by ANY of the following: RM-6535A-1 (Manipulator Crane) reading greater than 9500 m~lhr(Manipulator Crane) reading greater than 9500 mR/hr (ie Erratic source range monitor indication {4zWR}UNPLANNED increase in Kontainment Sumps A or (site-specifie zum e of sufficient magnitude to indicate core uncovery (Other zte ............. ticnVisual observation.Commented [DWS32]: V12 UFSAR Table 12.3-14 -CTMT 1 Post-LOCA Range [DWS33]: V13 Containment Sumps Basis: CONTAINMENT INTEGRITY: The procedurally defined conditions or actions taken to secure containment and its associated structures. systems, and components as a functional barrier to fission product release under shutdown conditions. UNPLANNED: A parameter 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'he known or unknown.46 This IC addresses a significant and prolonged loss of reactor vesseliRCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.Outage/shutdown contingency plans typically provide for re-establishing or verify'ing CONTAINMENT INTEGRITY following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs 1 .b and 2.b reflect the fact that with CONTAINMENT INTEGRITY established, there is a lower probability of a fission product release to the environment. In EAL 3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. Manipulator Crane setpoint of 9500 mRihr is 95% of the monitor range.The inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS. RVLIS LEVEL -VESSEL LEVEL (%) (inches from vessel____________ flange)~108 119.8 100 81.3 90 31.8 80 -17.7 70 I -67.2 63 -101.9 RC-Lt-9405, RC-LIT-9467, 60 -116.7 and the Tygon Tube do 55 -141.5 not indicate reactor vessel 50 -166.2 level when actual level is 40 -215.7 less than -95" due to the 30 -265.2 wetr on the RCP 20 -314.7 discharge. 10 -364.2 0 -413.7 _________These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal;SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG- 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Acttons to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CGI1 or RG I.47 CA1 ECL: Alert Initiating Condition: Loss of reactor vessel/RCS inventory. Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The Emergency DirectorSTED/SED should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. I (1) Loss of reactor vessel/RCS inventory as indicated by full range < 64% (-969")vc1 .. .....il-z specifi level. l"_ - [DWS34]: V1I EPCALC-06-04 -RVL1S Values OR (2) a. Reactor vessel/RCS level cannot be monitored for 15 minutes or longer.AND b. UNPLANNED increase in ......i zump. and/cr tan4 Sumps A or levels due to a loss of reactor vesselIRCS inventory. -{ Commented [DWS35]= v13 Containment Sumps Basis: UNPLANNED: A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or exent may be known or unknow~n.This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.For EAL # 1, a lowering of water level below 640%(zite sp,-,i:ic le-el) indicates that operator actions have not been successful in restoring and maintaining reactor vessel/RCS water level.The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.Although related, EAL # 1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., lcc c, fa Recidual Heat Reea ..... ti ... * .. t..:* An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.For EAL #2, the inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS. 48 RVLIS LEVEL (%) 100 90 80 70 64 60 50 40 30 20 10 0 VESSEL LEVEL (inches from vessel flange)119.8" 81.3" 31.8"-17.7"-67.2"-96.9"-116.7"-166.2"-215.7"-265.2"-314.7"-364.2"-413.7" RC-LI-9405. RC-LIT-9467. and the Tygon Tube do not indicate reactor vessel level when actual level is less than -95" due to the weir on the RCP discharge. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CSI If the reactor vessel/RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.49 CA2 ECL: Alert Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.Operating Mode Applicability: 5, 6, Defueled Emergency Action Levels: Note:...The Emrec ... roc......ST1ED/SED should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.l For a bus to be considered energized from SEPS. both SEPS die~sel generator sets must be functional. (1) Loss of ALL offsite and ALL onsite AC Power to BOTH AC emergency buses E5 AND.... (zt .......... cme.... busec)for 1 5 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.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service.Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.l-EDE-SWG-5 (E5) and I-EDE-SWG-6 (E6) are the 4.16 kV emergency buses for Train A and t'rain B? respectively. These buses supply all safety-related loads.[his Initiating Condition is not met if either Bus ES or E6 is energized from the Supplemental Emergency Power System (SEPS).The SEPS primary function is to supply power to one 4.16 kV emergency bus. EDE-SWG-5 (ES) or EDE-SWG-6 (E6), in the event of a loss-of-offs ite-power (LOOP) and both [D~s fail to start and load. In addition (SEPS) provides back up power to the emergency buses when one of the emergency diesel generators (EDO) is out of service for up to fourteen days. SEPS can be used when it is anticipated that one of the ED~s will be inoperable for longer than the technical specification allowable outage time (AOT) of 72 hours.The design of the SEPS is capable of providing the required safety-related loads in the event of a loss of offsite power if both emergency diesel generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (SI. CB3S, CVI, Cl, etc.). In addition to providing power to the required loads. the total combined output of the SEPS system can supply either the RHR pump or the SI pump and one set of pressurizer heaters. These design conditions are based on Probabilistic Risk Evaluation (PRA)[[-03-007. 50 The SEPS consists of two 4.16 kV generators which use diesel fuel engines as the prime mover.The generator sets (gensets) SEPS-DG-2-A and SEPS-DG-2-B are capable of automatically starting, synchronizing together and energizing the SEPS electrical bus. The SEPS design requires a 'dead bus" transfer back to an offsite power source, i.e.. the emergency bus powered by SEPS must be dc-energized before restoring offsite power.For power restoration from the SEPS, both SEPS diesel generator sets must be functional. The use of the SEPS is recognized in the Emergency Operating Procedures.

Reference:

VESAR Section 8.3.1, AC Power Systems Escalation of the emergency classification level would be via IC C S1 or RS 1.51 CA3 ECL: Alert Initiating Condition: Inability to maintain the plant in cold shutdown.Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The Dir......STED/SED should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely he exceeded.(1) UNPLANNED increase in RCS temperature to greater ...... e-seei;ie Technical Specificat.... col t..m. ra...r. limit)" for greater than the duration specified in the following table.---] Commentedl [DWS36]: V15 Cold SD Temp Limit TS Table4-( 1 -RCS Heat-up Duration Thresholds RCS StatusCONTAINMENT INTEGRITY Ha-pDrto RCS Status ~~Status Ha-pDrtoreactor vessel >- Not applicable 60 minutes*36 inches .*W4_____________________________ Not INTACT reactor vessel Established 20 minutes*mce .........Established 0 minutes* tf an: RCS he-at zyztzma.RlI R is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

Comete [ -S3] Vt eue net OR (2) UNPLANNED RCS pressure increase greater than psig reading-). (This EAL does not apply during water-solid plant conditions.)

Basis: tUNPLANNED: A parameter change or an event that is not I) the result of an intended ev.olution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknosxn.INTACT: Capable of being pressurized. This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT INTEGRITY is established but the RCS is not intact, or RCS inventory is reduced (e.g., midwv l....p 3per-at-,n in: The 20-minute criterion was included to allow time for operator action to address the temperature increase.The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT INTEGRITY is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-52 [DWS39]: Vt7 RCS Pressure range minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety.Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory [PWR], and CONTAINMENT INTEGRITY is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2)there is reduced reactor coolant inventory above the top of irradiated fuel.EAL #2 provides a pressure-based indication of RCS heat-up. The wide-range RCS pressure transmitters have a range of 0 to 3,000 psig. The main control boards have two post-accident monitoring qualified meters. one for each wide-range RCS pressure transmitter. ['hese meters have major divisions at 100 psig intervals and minor divisions at 50 psig intervals. Since it is possible to read the approximate mid-point between minor divisions, the value is set to 25 psig.Re ference: OS 1000.09 Escalation of the emergency classification level would be via IC CS 1 or RSl1.53 CA6 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Operating Mode Applicability: 5, 6 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: I. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.Basis: 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.EXPLOSION: A rapid. violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (fr'om 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 detenrmine if the attributes of an explosion are present.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. SAFETY SYSTEM: A system reqtuired for safe plant operation. cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. Systems classified as safety-related.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 safety of the plant.54 EAL 1 .b. 1 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. -EAL 1 .b.2 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 would be via IC CS1 or RS1.55 CUl ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of reactor vessel/RCS inventory for 15 minutes or longer.Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The ......nz Di... z......STIED!SED should declare the Unusual Event promptly upon I determining that 15 minutes has been exceeded, or will likely be exceeded.( I) UNPLANNED loss of reactor coolant results in reactor vessel/RCS level less than a required lower limit of an operating band, specified by an operating procedure for 15 minutes or longer.OR (4-)(2) a. Reactor vessel/RCS level cannot be monitored.-AND)b. UNPLANNED increase in (s:te ump ..... zr ........ontainment Sump A or levels.e. --(Commented [DWS40]: V13 Containment Sumnps Basis: UNPLANNED: A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor reactor vesselJRCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.Refueling evolutions that decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.EAL #1 recognizes that the minimum required reactor vessel/RCS level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.EAL #2 addresses a condition where all means to determine reactor vessel/RCS level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against 56 other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAl or CA3.57 CU2 ECL: Notification of Unusual Event Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer.Operating Mode Applicability: 5, 6, Defueled Emergency Action Levels: Notes:* The Emergency DirectorSTEID/SED should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.*For power restoration from the SEPS, both SEPS diesel generator sets must be functional. (1) a. AC power capability to Both AC emergency buses E5 ANI) E6'* oi v ......eme~enc bues)is reduced to a single power source for 15 minutes or longer.AND b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.Basis: SAFETY SYSTEM: A system required for safe plant operation, cooling down the ptant and/or placing it in the cold shutdown condition, including the ECCS. Systems classified as safety-related.This IC describes a significant degradation of offsite and onsite AC power sources 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. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service.Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition ~ presented below.A loss of all offsite power v,'ith a concu~ent failure of all but one emergency power source (~an oncite diesel generator). A loss of all offeite power and loss of all emergency power sources (e.g., onsite diesel P. loss 31 emergency power sources ~e.g., onsite dieset generators) wim a singte train or genemtors) with a single ~in of emergency buses being back fed from the unit main generalI I J wJ U *.1 m I *emerg..nc. buse being back;. from a.. .....t p* ower .ource.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.58 1-EDE-SWG-5 (E5) and l-EDE-SWG-6 (E6) are the 4.16 kV emergency buses for Train A and Train B respectively. These buses supply all safety-related loads.This Initiating Condition is not met if either Bus E5 or E6 is energized from the Supplemental Emergency Power System (SEPS).The SEPS primary function is to supply power to one 4.16 kV emergency bus. EDE-SWG-5 (ES) or EDE-SWG-6 (E6), in the event of a loss-of-off'site-power (LOOP) and both EDGs fail to start and load. In addition (SEPS) provides back up power to the emergency buses when one of the emergency diesel generators (EDG) is out of service for up to fourteen days. SEPS can be used when it is anticipated that one of the EDGs will be inoperable for longer than the technical specification allowable outage time (AOl) of 72 hours.The design of the SEPS is capable of providing the required safety-related loads in the event of a loss of offsite power if both emergency diesel generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (SI, CBS. CVI, Cl, etc.). In addition to providing power to the required loads, the total combined output of the SEPS system can supply either the RHR pump or the SI pump and one set of pressurizer heaters. These design conditions are based on Probabilistic Risk Evaluation (PRA)EE-03 -007.The SEPS consists of two 4.16 kV generators which use diesel fuel engines as the prime mover.The generator sets (gensets) SEPS-DG-2-A and SEPS-DG-2-B are capable of automatically starting, synchronizing together and energizing the SEPS electrical bus. The SEPS design requires a "dead bus" transfer back to an offsite power source, i.e., the emergency bus powered by SEPS must be dc-energized before restoring offsite power.For power restoration from the SEPS. both SEPS diesel generator sets must be functional. The use of the SEPS is recognized in the Emergency Operating Procedures.

Reference:

IJFSAR Section 8.3.1, AC Power Systems The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.59 CU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED increase in RCS temperature. Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The Emergenzy DirecterSTED!SED should declare the Unusual Event promptly upon I determining that 15 minutes has been exceeded, or will likely be exceeded. I (1) UNPLANNED increase in RC S temperature to greater than --( Commented [DWS41 ]: v15 cold SD Temp Limit TS OR (2) Loss of ALL RCS temperature and reactor vessel/RCS level indication for 15 minutes or longer.Basis: LUNPLANNEID: A parameter 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 evcnt may he knowvn or unknown.This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT INTEGRITY is not established during this event, the Emiergeshey DieeS rED/SEID should also refer to IC CA3.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. EAL #1 involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. 60 Escalation to Alert would be via IC CA! based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.61 CU4 ECL: Notification of Unusual Event Initiating Condition: Loss of Vital DC power for 15 minutes or longer.Operating Mode Applicability: 5, 6 Emergency Action Levels: Nt:The E..r... z Dir.. ec.....STED/SED should declare the Unusual Event promptly upon I determining that 15 minutes has been exceeded, or will likely be exceeded.I ( 1) Indicated voltage is less 1 05 Vite zpecific vzltag¢ "alue) on required Vital DC ---[ Commented [DWS42]: Vis UFSAR 8.3 2- DCV 105 limit buses associated with the Protected Vrain for 15 minutes or longer.Train A 11A and I1C Train 13 1113 and l ID Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode.In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Per DBD-ED-05. the DC bus voltage range within w~hich the 125 Volt DC system is considered operable is 105 volts minimum to t40 volts maximum. The vital DC Buses (Switchgear) are SWG-t1IA and I IC lor Train A and SWG-l1IB and I ID for Train B.

Reference:

UFSAR Section 8.3.2. IDC Power Systenm Procedure OS 1248.01. Loss of a Vital 125 VDC Bus Procedure VPRO ['5278. Loss of All Vital I)C Power DBD-ED-05, 125 VDC Systenm Depending upon the event, escalation of the emergency classification level would be via IC CAl1 or CA3, or an IC in Recognition Category R.62 CU5 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities. Operating Mode Applicability: 5, 6, Defueled Emergency Action Levels: (1 or 2 or 3)(I) Loss of ALL of the following onsite communication methods: In-Plant (PBX) Telephones Gai-Tronics Plant Radio System OR (2) Loss of ALL of the following ORO communications methods: Nuclear Alert System (NAS)Backup NAS All plant telephones Cellular telephones OR (3) (ste .p.. i.. lict ,of co, ....... on, .....h^,ds) Loss of ALL of communications methods: Emergency Notification System (ENS)All plant telephones FTS telephones in the TSC Cellular telephones [the following NRC 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., .... cf^ non lan, pr....... 3we equipm...nt, r.l..n ef .- cit 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 Commonwealth of Massachusetts and State of New Hampshire. 63 EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. 64 ATTACHMENT 2 Markup of Seabrook Station Emergency Action Levels -Initiating Conditions, Threshold Values and Basis SEABROOK STATION EMERGENCY ACTION LEVELS INITIATING CONDITIONS, THRESHOLD VALUES AND BASIS TABLE OF CONTENTS 1 REGULATORY BACKGROUND ..............................................................

1.1 OPERATING

REACTORS ........................................................................... 1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI).......................... 1 1.3 NRC ORDER EA-12-051 .................................................................... 2 1.4 ORGANIZATION AND) PRESENTATION OF INFORMATION................................. 12 1.5 IC AND EAL MODE APPLICABILITY........................................................ 13 2 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS............................ 21 2.1 GENERAL CONSIDERATIONS ................................................................. 21 2.2 CLASSIFICATION METHODOLOGY ............................................................. 22 2.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITIONS .............................. 22 2.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION ...................... 22 2.5 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING ............ 23 2.6 CLASSIFICATION OF SHORT-L1YED EVENTS............................................... 24 2.7 CLASSIFICATION OF TRANSIENT CONDITIONS .............................................. 24 2.8 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION........25 3 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS .................. 26 4 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS ............... 42 5 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS .......... 65 6 FISSION PRODUCT BARRIER ICS/EALS ................................................ 67 7 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ......51 S SYSTEM MALFUNCTION ICS/EALS..................................................... 104 APPENDIX A -ACRONYMS AND ABBREVIATIONS ........................................... 1 APPENDIX B -DEFINITIONS..................................................................... 4 ii

OF EMERGENCY ACTION LEVELS 1 REGULATORY BACKGROUND

1.1 OPERATING

REACTORS Title 10, Code of Federal Regulations (CFR), Energy, contains the U.S. Nuclear Regulatory Commission (NRC) regulations that apply to nuclear power facilities. Several of these regulations govern various aspects of an emergency classification scheme. A review of the relevant sections listed below will aid the reader in understanding the key terminology provided in Section 3.0 of this document.* 10 CFR § 50.47(a)(1)(i)

10 CFR § 50.47(b)(4)
10 CFR § 50.54(q)* 10 CFR § 50.72(a)* 10 CFR § 50, Appendix E, IV.B, Assessment Actions* 10 CFR § 50, Appendix E, IV.C, Activation of Emergency Organization The above regulations are supplemented by various regulatory guidance documents.

Three documents of particular relevance to NEI 99-01 are:* NUREG-O654iFEMA-REP-lI, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nuclear Power Plants]* NUREG- 1022, Event Reporting Guidelines JO CFR § 50. 72 and § 50. 73* Regulatory Guide 1.101, Emergency Response Planning and Preparedness for Nuclear Power Reactors Emergency Preparedness staff.1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)Selected guidance in NEI 99-01 is applicable to licensees electing to use their 10 CFR 50 emergency plan to fulfill the requirements of 10 CFR 72.32 for a stand-alone ISFSI. The emergency classification levels applicable to an ISFSI are consistent with the requirements of 10 CFR § 50 and the guidance in NUREG 0654!FEMA-REP-1. The initiating conditions germane to a 10 CFR § 72.32 emergency plan (as described in NUREG-1567) are subsumed within the classification scheme for a 10 CFR § 50.47 emergency plan.The generic ICs and EALs for an ISFSI are presented in Section g5, ISFSI ICs/EALs. IC E-44U-1E U I covers the spectrum of credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to !.inslations. or facilities tha my pocss n4or epckae pen fie (eg. a sntord et~vale toag The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-l1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees. NUJREG-1 140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the maximum offsite dose to a member of the public due to an accidental release of radioactive materials would not exceed 1 remn Effective Dose Equivalent. re~uirea Icr IU t.A-K ~ ~ emergency pian Is ornerent man mat presermep ror a mu CFR ~ 50.17 emeraenc'i plan (e.a.. no emergency technical support funetien). 1.3 NRC ORDER EA-12-051 The Fukushima Daiichi accident of March 11, 2012, was the result of a tsunami that exceeded the plant's design basis and flooded the site's emergency electrical power supplies and distribution systems. This caused an extended loss of power that severely compromised the key safety fractions of core cooling and containment integrity, and ultimately led to core damage in three reactors. While the loss of power also impaired the spent fuel pool cooling function, sufficient water inventory was maintained in the pools to preclude fuel damage from the loss of cooling.Following a review of the Fukushima Daiichi accident, the NRC concluded that several measures were necessary to ensure adequate protection of public health and safety under the provisions of the backfit rule, 10 CFR 50.109(a)(4)(ii). Among them was to provide each spent fuel pool with reliable level instrumentation to significantly enhance the ability of key decision-makers to allocate resources effectively following a beyond design basis event, To this end, the NRC issued Order EA- 12-051, Issuance of Order to Modify Licenses with Regard to Reliable Spent Fuel Pool Instrumentation, on March 12, 2012, to all US nuclear plants with an operating license, construction permit, or combined construction and operating license.NRC Order EA-1 2-051 states, in part, "All licensees ... shall have a reliable indication of the water level in associated spent fuel storage pools capable of supporting identification of the following pool water level conditions by trained personnel: (1) level that is adequate to support operation of the normal fuel pool cooling system, (2) level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck, and (3) level where fuel remains covered and actions to implement make-up water addition should no longer be deferred." To this end, all licensees must provide:* A primary and back-up level instrument that will monitor water level from the normal level to the top of the used fuel rack in the pool;2

A display in an area accessible following a severe event; and* Independent electrical power to each instrument channel and provide an alternate remote power connection capability.

NEI 12-02, Industry Guidance for Compliance with NRC Order EA-12-051, "To Modify~Licenses with Regard to Reliable Spent Fuel Pool Instrumentation ", provides guidance for complying with NRC Order EA- 12-051.NEI 99-01, Revision 6, includes three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order BA- 12-051. These EALs are included within existing IC RA2, and new ICs RS2 and RG2. Associated EAL notes, bases and developer notes are also provided.It is recommended that these EALs be implemented when the enhanced spent fuel poo1 level instrumentation is available for use.The regulatory process that licensees follow to make changes to their emergency plan, including non-scheme changes to EALs, is 10 CFR 50.54(q). In accordance with this regulation, licensees are responsible for evaluating a proposed change and determining whether or not it results in a reduction in the effectiveness of the plan. As a result of the licensee's determination, the licensee will either make the change or submit it to the NRC for prior review and approval in accordance with 10 CFR 50.90.3 2nEYv TERMunOLOnnGY nnaED In N-"n BO_^There are ceveral key term:n~ that appear throug~hout t.he NEL 99 01 methodology. These term":. a.e i-ntroduced': in hi:. ceetion toc upp.ort of :ube.guent material2. A: an aid to the reader, follow--ing table ic provide~d a: an ove'r.'i-,- to th'e relatincnhip c~f the Emergency Em...erge..cy Emergency Emergency'.Notes *--Nte --Notes '-Note:*---Basi 2.1 EMERCENCY CLAn~LrICATIoN LEVEL (ECU One of a Oct of name: or title: cotabliohed by the US Nuclear Regulatory Commi::ion (NRC) for grouping off normal event: or condition: according to (I) potential or aemal effect: or conoequonce:, and (2) reculting on:ite and offoite re:pon:e action:. The emergency ela::ification level:, in acending order of se~ ~rit~, ar~.U Notification of Unu:ual Event (NOUE)Ul-A-lei* itc. A~rea (^AE)£ General Emergency (GE)I -I ~TITIanhIvn flT I inUIIflI ,-.venr Ira fl I-. I lee o ... f safet" of the*. plant or indiate.... a......o.

ecurit threat to. facili.t" protection has been.*expected unle:: fu"rther degadation of :afct s~tm ocu" 4 Pu:rpesi:

T he purpoce ot tux: s classification is to assure that 1 t, to bring the operations staff to~e4~e0 4ii-fu~e a state of readiness, and to.~2I~-A4ei4 Events are in progress or have occu..rred -which involve an act+'ual or poten:tial s.botantial degradation cf the l....l of cafc, of +the -plant cr a c'eeu-' eve... tha ... pr...obatle plfe threatenin reede tort peretcnnen efrh puagc to HOSTeqiLmEn beAuCTO ofa rOuT iLE ACTON.Anyrelasc ar exectd t b liite toomal faetoncof he EPA PA...e .. ..... ...e .... .. .... .. .. ..... ... neplu e__;_Pupose: Te pup ose= + of' thic: claci..... o i c t....... asur ta eer-cypesnnl r 21. I General Emerzencv ,(GE), Evento are i:n progreoc or have occu"-red -which i.nvolve actual: or su.'bstantial ACTION that recul:to in an actual !boo of phycical control of the facili.ty. Releases can be reaoenabkl, expected to exceed EPA DPAGP expocure. leve+lsoffcit for more tha n the+the,.,ou li...... and offcite organizational measurement, to..initiate.additional.me..ures.as 5

2.2 INITIATIiNC

Co~inITLoN (IC)An event or condition that aligns with the definition of one of the four emergency clazoifloation levelo by virtue of the potential or actual effecto or consequences. Discussion: An IC describes an event or condition, the severity or consequences of which meets the definition of an emergency classification level. An IC can be expressed no a continuous, measurable parameter (e.g., RCS leakage), an ovent (e.g., an earthquake) or the status of one or more fission product barriers (e.g., boo of the RCS Appendix I cfNUREG 0651 doeo nct contain example Emergency Action Levelo (EAIo) for each ECL, but rather Initiating Conditions (i.e., plant conditiono that indicate that a radiological emergency, or evento that could lead to a radiological emergency, hao occurred). NUREG 0651 otateo that the Initiating Conditions form the baoio fc~eotablio~ent by a lieenoee of the specific plant ino~mentation readingo (go applicable) which, if cxceede~ would initiate the emergency clasoification. Thus, it io the specific inotrument readingo that would be the EALs.Considerations for the asoignment of a p~icular Initiating Condition to an emergency clasoification level are discusoed in Section 3.2.3 E~ncLNcY AC~ON LE~L (EAL)A pre determined, oite specific, oboer;ablc threshold for an Initiating Condition that, when met or exceeded, places the plant in a givcn emergency claosification level.Discussion: EAL otatemento may utilize a v~ety of criteria including inotrament readingo and otatuo indications; oboervable events; resulto of calculationo and analyocs;entry into particular procedureo; and the occurrence of natural phenomena.

2.1 FIsSIoN

PnoDucrr BARmEn TIIUESIIOLD A pre determined, oite opecific, obser.'able tI potential boo of a fission product b~cr.~rcofloIC meicating h los or Discussion: Fiooion product barrier threoholdo represent threats to the deI~nse in depth design concept that precludes the relenoc of radioactive fiooion producto to the environment. Thio concept relieo on multiple phyoieal barriero, anyone of which, it maintaincd intact, preoludeo the rcleaoe of oigniflcant amounto of radioactive fiooion produeto to the environment. The primary flooion product barriers arc: E-4**el-Gla4 U Reactor Coolant System (RCS)E-Ce~ai~en~ Upon determination that one or more fiosion product barrier threoholdo have been sxcs~de& the o~mhination of barrier boo and/or notential boo thrcoholdo is zomoared to the fiooion product b~ier IG~,AI criteria to detcrmi In some accident oequences, the ICo and EAt 6 ne thc appropriate ECL presented in the Abnc~mal Radiaticn L~eIci Radizlogizal Effiuznt (A) Rc~cgniti3n Categc~' ~'ill b~ zxcedad at th~ znm~ time, cr shortly after, th~ loz of on~ or mzrz fission product barri~r~. This r~dundanzy is intantftmal as thz fcrmer ICs addrzss mdioasti~'i~' rzIc~ses that rzsult in zzrtain zffsit: d~s~s fram whztz;zr zauoc, including events that might net be fully cnocmpassed by fissicn preduct barriers (e.g., spent fuel pool accidents, design centainment leakage fcllowing a LOCA, etc.).7 A-3 OFl lTrhi NEDl a990_1 EMEl~rGENCrY d~l CA~g25ItICTIONmlk SCrH~EME h 3.1 A:SICN1~U~NT or EMsncENv~' CLASSIrICATION LEVELS (ECLS)An effective emcraencv clacificaticn ceheme muct incorcorate a realictic and g1~iirnt~ n e~ment '~f rPdc h~th t'~ r.lnnt ~prkerc and th: hea~h-~~-sa~et~-rick: in Underectimating the pe er, there are alco rickc in overe: st-aiekg public. There are obviou: actual threat from an event cr the threat as well (e.g., harm condition; hcwcv: attempt: to ctrke an, a~pproprate betw;een reac.enably a:ntici:pated even~t ci There are a .-nge of "non emergency ev'ente" reported to the US N-o-le=" to determi..ne t.he attribute: of each ECL. The goal o~f thi: pro~s is tc an w'.er fc,,ll+'ow'ng

ource:

infc..-m.ation and c..nte.t f.,or the de..elopment. of ECL U Typical abnorm'=al and emergency operatng procedure etpoints and criteria U Typical Techn:ical Spcifcaton^ ,liio

Reniew ofelecedus sUpdaedt FialSer ' ........ s Report (F+o acidnt.nalc The follo'wing ECL attributes

-,e*re cerated by the Re-vicion 6 Pre:paration Team. to aid in include the attrbut:,.. in thi: ....io ince they' may be .......in.brie..n an training.. in thic docu'ment. fle attribute: or eaefl LLL ore precentec tetow.8 3.1.1 344- Notification of Unusual E vent O.TOUE) o-r unuIsua, t Lven, aS o.eiineo in secixon /. i. , o-.ur lo ,ni9-, :iimne.u L an event or condition tha involves: (A)A precursor to a more significant event Zr condition.(C) A consequence oth"r....e si agn~ificant enough t ......a..t notifcatio ... lal, tte an...d An Alert, as defined in section 2.1!.2, incluades bu-t is not limi~ted to an event or codiio ..a. inv...............l.......e.s:~tn a ia ,t control radiation levels w.ithin the plant, or a release of r.adioactive materia's to the 3.1.3 Site b=ea Emergency A Site Area Emergency, as defined in section 2.1!.3, incluades buat is not limited to ( A)"A laos or potential1 loss ofe+ any, product+ barriers c1ad, B)A peuro event....or conditi that, may. lead to the los or potentila loss of multiple , issio product barr..... withina.relatively..... period.of.tim ..Pecrsrvetsan conitsions --ofa this- ...pe include *°:"' those tha challeng themoitoin .......or .conto ot than.......f.an.EPA.P ...at or beyon..th site bounday 1 ACg'TION%. occu,,in within the.l- plant PRO+1D AEA 9

3.1.1 General

Emergency A General Emergency, ac defined in section 2. 1.1, includes bat is not limited to an, event or condition that in;'vels'e: pr.u.tba... r... Precur.or....ent..... conditions ofrtis- tpe include those that lead dirctly to c.. re. damage.. an loss of cortaimer.nt tha an. EP A ^P.G at or beyond the site boudar..4, (D)A HOSTILE ACTION rec.ulting in the o, ..f key ...... funtion..

(react..i..

-.,......., 3.1.5 Pick Info.rm.ed Insights i~mergency prepareJnesc xc a ~ietence in depth meacure tflat Ic nuependent 01 tile accessed rick from any particular accident sequence; ho;vever, the development of an effective emergency clascification scheme can benefit from a review of risk based acccccment recultc. To that end, the development and assignment cf certain ICc and EALs alco concidered incightc from several cite cpecific probabilictic cafety acccscmentc (PSA alco IGiown as probobilictic risk acceccment, PILA4 Seine generic incightc from thic review included: 1. A ~xd~nt '~guences involving a prolonged less of all AC power are significant ccntributorc to core damage frequency at many Preccurized Water Reactorc (PWRc)and Boiling Water Reactorc (BWRc). For this reason, a locc of all AC power for greater than 15 minutec, with the plant at or above Hot Shutdown, was assigned an ECL of Site Area Emergency. Precurcor eventc to a loss of all AC power were alco included as an Unucual Event and an Alert.A ctation blackout coping analyses performed in responce to 10 CFR § 50.63 and Regulatory Guide 1.155, Station B!aekoot, may be used to determine a time baced criterion to demarcate between a Site Area Emergency and a General Emergency. The time dimension is critical to a properly anticipatorj emergency declaration since the goal ic to minimize the time available for State and local officials to develop and implement offoite protective actions.2. For severe core damage events, uncertaintiec exist in phenomena important to accident pro~ecsionc leading to containment failure. Because of thece uncertaintiec, predicting the statas of containment inte~ity may be difficult und re ra ident conditianc. Thic ic why maintaining containment integrity alone following sequences leading to cevere core damage ic an insufficient basic for net escalating to a General~e~gei~ey~

3. PSAc indicated that leading contributorc tc latent fatalitiec were sequeneec involving a containment bypooc, a large Losc of Coolant Accident (LOCA) with early 10 e-.nt boozed ICs and Each type is discussed below.Symptom haz..ed ICs anrd E.A~sar"e or cond.itions are mea,.urable radiologieal effluent, etc.). When oneeor more of these param-eters or conditions, are off EAts that refer specifically to the level of challenge to the principal barriers against the rees of.... radioa.ti...mate

..al ..r.m.th.r.actor.core.to.the.environ ent.Te.ebrir ar- th"e'e+ fel ladnteractoa .r. .. coolan system a presr boudar, and.*... the.+ cont.inment Th...b....er.b...ed .....and .......consider ..the.le.el of ...hall ...e to each iniida br n" ...... 1 ....m~aoe nazar-as suen as a gas release.3.3 NSSS DESIGN DwrniE~clss The NE' 99 01 emergency. classificatio schem accounts... for the design differences betw;een PWP~s and BW?.-s by ....~,in EA.. s uni...ue to each... typ ci Dev'elopers --ill need to consider the relev'ant oopeets of their plant's design and site specific classification scheme. Th~e goal is to. maintai as much..... fi-lt as v.......The guidance in Nrct 99 01 is not applcable to. advanceed pasve light.. water...de vac'-eloped" i... aodac with.. NEIxr 0"7 01 Afthdo -g fr D.... op...... fE.'........ II 341.4 ORGANIZATION AND PRESENTATION OF GNIAiqEINFORMATION The scheme's-eei information is organized by Recognition Category in the following order.* R -Abnormal Radiation Levels / Radiological Effluent S4eedier. 5* C -Cold Shutdown / Refueling System Malfunction -Seectecn-7

E -Independent Spent Fuel Storage Installation (ISFSI) -4eletion8
F -Fission Product Barrier -Seetion-9
H -Hazards and Other Conditions Affecting Plant Safety-8eSeft-4n 1* S-M -System Malfunction S4ectionA4 The following inf..rmat...n..n.

guid..n...i. prvid..d........h..C. ECL the a:signecd emergency classification level fcr the IC.Operating Mded Applicabili-. Lit t ..he modes during which the IC and-If',- th ... g-en-ri' to the_. development of an example EPI cannot. b...e+ used (e.g., -an...umed instrumentation r...g. is net a.a.a. l the plat) the'* should .ttempt degee of bm'ier challenge (i.e., potentia l:ss ,or Thi.s presentation method shows..+the syn'ergism among tetrsodadspot cuaeassmns ofte ICan EA,^ s. In Tsom e ae, h assa includesea~ia~ ......... rlvan sour. in............ calculations, etc... D vepr n.... shold. nt...incudedin..e.ste.semerenc classification scheme basis Developers may elect to include inform,.ation £ ECL A~ssignmen~t Aftr'ib.tcs Lccate:d w,:ithin the Developer Notes scecti12 3J.1.5 IC AND EAL MODE APPLICABILITY The NEI 99 01 emergency classification scheme was developed recognizing that the applicability of ICe and EALs will vary with plant mcde. For example, seine symptom hazed ICe and EALs con be assessed only during the power operations, startup, or hot standby/shutdcwn modes of operation when all fission product barriers ore in place, and plant instrumentation and safety systems are fully operational. In the cold shutdown and refueling modes, different symptom booed ICe and EALs will come into play to reflect the opening of systems fcr routine maintenance, the unavailability of some safety system components and the use of alternate instrumentation. The following table shows which Recognition Categories are applicable in each plant mode. The ICs and EALs for a given Recognition Category are applicable in the indicated modes.MODE APPLICABILITY MATRIX Category Mode R C E F H SM Power Operations X X X X X Startup X X X X X Hot Standby X X X X X Hot Shutdown X X X X X Cold Shutdown X X X X Refueling X X X X __Defueled X X X X 13 Operating Modes Technical TABLE 1.2-. --COmmented [DWS1]: VI TS Table 1.2 Mode Defleition MODE Reactivity % Rated Thermal Average Coolant Condition. Keff Power* Temperature

1. Power Operation 2. Startup 3. Hot Standby 4. Hot Shutdown 5. Cold Shutdown 6. Refuleling**

NA Defueled> 0.99> 0.99< 0.99< 0.99< 0.99 NA> 50%< 5%0 0 0 0_> 350 0 F> 350°F_> 350°F 350 0 F > Tavg >200 0 F< 200 0 F< 140 °F All fuel removed from thle reactor vessel (full core offload during refueling or extended outage)*Excluding decay heat.**Fuel in the reactor vessel with the vessel head closure bolts less than fully' tensioned or wvith the head removed.Pe r Operat en (1) Reacte ...... r 5.... Koff ¢-Strup (2):*, ...a. to, ",. wer 5t%, Ke-- 1.6StadbB(3:ISS 35OCUMENT- .9 proidnginfrmngbakgrun adRevuelipmngt6) Onento in headil acelsuib forma. Itcan e reerre to n bri ltn lsiutos tand whlyen maionednatleerec cle asisiiainifecsa.Th document is anlnerasat ofa emerencu larssabifhic gcniuationsce .Th mnanagemaent controls for EP-related equipment and explaining an emergency classification to offsite authorities. The content of the basis document includes:*A site-specific Mode Applicability Matrix and description of operating modes (see Section 1.5).14

A discussion of the emergency classification and declaration process (see Section 2).* Each Initiating Condition along with the associated EALs or fission product barrier thresholds, Operating Mode Applicability.

Notes and Basis information (see Sections 3-8).* A listing of acronyms and defined terns (see Appendices A and B. respectively). A basis section should not contain information that could modify the meaning or intent of the associated IC or EAL. Such information should be incorporated within the IC or EAL statements, or as an EAL Note. Information in the Basis should only clarify and inform decision-making for an emergency classification. Basis information should be readily available to he referenced, if necessary. by the Short Term Emergency Director/Site Emergency Director (STED/SED). For example, a copy of the basis document could be maintained in the appropriate emergency response facilities. Because the information in a basis document can affect emergency classification decision-making (e.g.. the STED/SED refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q).I1.7 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA The criteria/values used in several EALs and fission product barrier thresholds may be drawn from AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments. Appropriate administrative controls are in place to ensure that a subsequent change to an AOP or EOP is screened to deternine if an evaluation pursuant to 10 CFR 50.54(q) is required.15 I SITE-SPECIFIC SCHEME DEVELOPMENT GUIDANCE This ecti. prvdsCeale uncep~z!,Leapoc fr d....l.ping ... sit .p.. ii .. m...rg...ny .l...iicat.. ....r g ..n ..y -pr t n p r o. .d.r. g en.. .r i. m at. r ia l p r e p a d b y ......... .,en d o o w ;n e r g r o u p s is c o n v e rte d b y e a c h ...u l r p o.. .r.. l. n t i n to: ...: ...t .sp e ifi ..m .rg e ... *pr t n r e d r Likewise, the emergency classificatio ..h.. develop.. r I .. ll. us1 ..h e generic guidanc ... NE TIT It is important that the NEI 99 01 emergency classification scheme be i'mplemented as an integrated package.. Slcted u.. se.. of portions of this guidanceo is strongly, disc...ur.ged4 as it ....l l.ad to. an in.n...en or4- incompl..

t. eme...rgency classificatio

..h. m tha will. l.;11ikely net r...eiv.4.1 CENEnAL IMPLEMENTATION CUIDANCE The g..uidance. in NE! 99 01 is n..t intendedn to b. .pplie tn plats+ ,as is";.generic guidance as p.sib.. Th li ometteitn ..f. th generic..n.t.a..ng chaPracteIstics. l.cale, plan desima, operati-ng features, te....i.-noloy et. Meetig thi ge lwll rslt in a sherter ,andlaa4 h les ubro rac, NRC rev++h.iew and approval ..... e,4 a,: o,,sessahle. As discu...sed in Section 3, the generi guidance:, incl:udes,4o Ic and E ^' °se.'ves a. spec7ific prpose...The....i..the.f.nd...ntal.e.ent er condition.. requi..ng c If++ some. featu+...re" of.. .........o .. or d ..sign. is+ ne comptibl with a generic IC er If-,an IC or EPJL incluades explicit reference to a mode dependent tec!=ieal spciiation that is not applicable to the plan÷ then- t+hat ICm an-,4In E2AI nee .net be hnclued inhe sitespecific schnteme. InThese worsc -dtevelpes mustproidentfe ad equatset clsification, ..heme (i re. LLCPr S).nTed 4.::........ in Appendix AAcemsadbreitiosSt 16 precentaticn pu:rpoece (e.g., "with a General Emergency at the* left'top of a user aid, follewed by Site Area Em..rg..n.y, .l.rt an NOUE).* M may be 'used in lieu- ef S* The IC: and EAL:s from Recognition Categoriec S .anC ma be. inorpor... ted int :..common precentation method (e.g., ....e .ale provided.that.al..related.n.t..... £ The. and EAL, for Emergency Direetor judgument an .....it relat.ed e..ent. may£ The term:n EAL and th~rechold may be used interchangeably. clascificatio scheme basics document.4..... TtcAL Cnni. .canisT'c charactericstics arc !ited below.£ The IC:, EAL:, Operatin Applica':'bilt citeri,% Note: n Ba- i information..:^ U hEA IC:, EALnt Operajeting"oe Arpplicabilid criter','~ Notles.n accifomt!17 i;;'hnere nem tc iltts prog n owgaig ttl"y.gnc lsitcto U z. The facilitates classification of multpe ...n.u...nt....nt. or condfitins, 1.3 INSTflUMENTAflON UCED ron EAL~,n tr... ntation ref.r. n e in.. EAI: s^tat..m.nt. should include described in the Technical, pe.cification or DMIRTScontr,.l requirement, nor po'wered from an Scheme. devl.. r , should ensur.... tato .p..fi. value"- .. used as EA ^s etpcints are -"ithin t.he calibrated range of the referenced ins~amcn~tatn, and cons~ider any automatic and v'iolations related to, EAL instrmnentation issues may be located on the NPRC -website.1.1 PREaENTATION or SChEME INrOIIMATtON TO U~Ens within: 15 minutes' afte the aviablt of indication~s tocln prtr hta pre.ent.tio method. s bet. ..upp..... theen users by facilitating accurate and timely.......... classification Tc this end, developers should conide the follow-ing points....Th..fi ..t.use ...of.an emegec cl.. si...cation. procedure are the oeramtor, in the.respnsibli. to,:. perf.rm other. critical tasks , and. wil likeyhav minimal..... assstan:eo in makldng a classification assessment.

,^ an emergency situation ev-les-e, membes of the Control Rooem staff are likely to* Emergency Directors in the TSC and/or EOF -will have more opportunity te' foe'-u on m.. ing anemrenycl.ifctin and........................probably...ha.e a visrs frmOer.atin avaiableto elp hem 18 Diretctrs and.'or Offsite R~esponse Organization personnel.

A "A,vubu°.d is an a:em~ presenmtatn me.ke pro;'.'ded +" tcz*n all+e..nclu..:..h...C.. Operat.ng Mo..... ppli..ab.l.ty .......a, ......d. ot....N.. t... may be wallboard bu-t it shoul:d be readily to emergentcy clas.sification decisica maker:.AtRnaie prsn Cation. method: fcr the Rec + gnition Cat... g..... F I:, .ad fis.ion... prouc to promote consistenc~ across mc Inoustry.1.5 I!mcRATION' or IC°JEALg ".rrTn PLANT PRC"CED'JRES A rigorous integration of IC and EAL references into plant operating procedures is not recommended. This approach would greatly increase the adminis~ive controls and workload for maintaining procedures. On the other hand, pefformance challenges may occur if recognition of meeting an IC sr EAI is based solely on the memory of a licensed operator or an Emergency Director, especially during periods of high stress.Developers should consider placing appropriate visual cues (e.g., a step, note, caution, etc.) in plant procedures alerting the reader/user to consult ~e site emergency classificatisn procedure. Visual cues could be placed in emergency operating procedures, abnormal operating procedures, alarm response procedures, and normal operating procedures that apply to cold shutdown and refueling modes. As an example, a step, note or caufion could be placed at the beginning of an RCS leak abnormal operating procedure that remind: the reader that an emergency classification assessment should be eifei~ied 7 4.6 BAsIS DOCUMENT A bais* doumnti an inera arfaneerec elassitication scheme. the matcr:al in tnin aecument supports proper emergency ciassuncanoen aeesion ma~rng ny provucin:g can orefcre to in' training situation and. whn m., king. an" actu'al emergency 19 a minimum, the following:

A site cpecific Mode App'icability Max a znd decoription of operatingcimilar to mat precentee in ceenon iz.£ A diccuccion of the emergency elaccification and declaration preceec reflecting the material precented in Section 5. Thic material may be edited ac needed tc align with cite epecifie emergency plan and implementing procedure reguirementc.

£ Each Initiating Condition along with the accociated EALe or ficcion product barricr thrcchcldc, Operating Mode Applicability, Notoc and Bacic informatien. £ A licting of acrenyme and defined termc, cimilar to that presented in Append~cec A and B, respectively. Thic material may be edited as necded to align with cite speeifie eha~eteristie~

~ specific background or tcchnical appendices that thc devLp~r.~bJi~...

weuld be useful in explaining or using element of thc emergcncy classification sehe~e A Basic section should not contain information that could modify the meaning or intent of the acceciated IC or EAt.. Such information chould bc incorporated within thc IC ot LAd. statements, or as an EAL Note. Infcrmation in the Basic should only clarify and inform decision making fcr an emergency classification. Bacis informatien should be readily available to bc referenced, if necessary, by the Emergency Dfreetor. For exumple, a copy of the basis document could be malntair.ed in the appropriate emergency response facilities. Because the information in a basis document can affect emergency elacsifieation decision making (e.g., the Emergency Director refers to it during an event), the NRC staff expect that changes to the basic document will be evaluated in accordance with the provicionc ef 10 CFR 50.51(q).1.7 EAL'TtrnESIIOLD RErERENCLIS TO AOP AND EOP SETPOINT~ICIUTEJ1IA Ac reflected in the generic guidance, the criteriaivalues used in several EALs and ficcion product barrier thresholds may be drawn from a plant's AOPc and EOPc. This approach Ic intended to maintain good alignment between operaticnal diagnoces and emergency elacoification assescmentc. Developere sheuld verify that appropriate administrative controls are in place to encure that a subsequent change to an AOP or EOP ic cereened to determine if an evaluation purcuant to 10 CFR 50.51(q) ic required.1.8 DEVELOPEB AND UTEEn FEEDBACK or comments concernring tne material in........... ay....... to t tn NEL merenc Prpardnec s, NIU .......force ......r. or cubmired~g to the ETmer c--ar Pea dns rq nty Acked Que......nc ....... ........ ..... ...20 62 GUIDANCE ON MAKING EMERGENCY CLASSIFICATIONS 54-2.1I GENERAL CONSIDERA&TIONS When making an emergency classification, the Emergencey DirectorSTED/SED must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes and the informing Basis information. In the Recognition Category F matrices, EALs are referred to as Fission Product Barrier Thresholds; the thresholds serve the same function as an EAL.NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, Interim Staff Guidance, Emergency Planning for Nuclear Power Plants.All emergency classification assessments should be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding the indicator's operability, the condition's existence, or the report's accuracy. For example, validation could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel. The validation of indications should be completed in a manner that supports timely emergency declaration. For ICs and EALs that have a stipulated time duration (e.g.,* 15 mi.n...,,,: 30.. minutes,,, t..)...the Emergency DirecterSFEI)/SEI) should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EAL has been exceeded, absent data to the contrary.A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that l) the activity proceeds as planned and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that compliance is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements ofl10 § CFR 50.72.The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g., dece ........... chemict.r' zampling, RCS !cak rate calculatien, etc.); the EAL and/or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC 21 expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., the. ........ , ' en ehift).While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 scheme provides the Emergneny DireeturSTED/SEI) with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emer~gencey will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated into the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.&2.2 CLASSIFICATION METHODOLOGY To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL(s) must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, then the IC is considered met and the associated ECL is declared in accordance with plant procedures. When assessing an EAL that specifies a time duration for the off-normal condition, the"clock" for the EAL time duration runs concurrently with the emergency classification process "clock." For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01.$---32.3 CLASSIFICATION OF MULTIPLE EwENTs AND CONDITIONS When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared.For example:* If an Alert EAL and a Site Area Emergency EAL are met, whether at e~ne u.nit e~r at..t, ..different uni,,,* a Site Area Emergency should be declared.There is no "additive" effect from multiple EALs meeting the same ECL. For example:* If two Alert EALs are met, whether at..... uni er at... twe .iffer.n .uni an Alert should be declared.Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events.CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is 22 declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original exent or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition. [or events that occur in Cold Shutdown or Refueling, escalation is via EAI~s that are applicable in the Cold Shutdown or Reftteling modes, even if H~ot Shutdown (or a higher mode) is entered during the subsequent plant response. Inl particular. the fission product harrier EALs are applicable only to events that initiate in the Ilot Shutdown mode or higher.Onemadiffaernt mo dei ece. n e events or condition, nhat ruldated to thein oriecedin n AL withi a renditiven, shrtquiring oemergien., lasfchatgein hudbe ECaLuatsdIMgINENT)th If, i and jugent aolialft the Eperaeing mrcderatthE/SD tmeftetinew avntEA ir IMIENTiton apliabergincy clasfctonsold beto~o madueln mass c if heELHasote meut.o (hir appligabe mode ll enterged curingfthatiubseeuentplant rppponaeh In particularly thpofisntitntherhduct barrergency carssaipicabinlevenly tsivnte tht poineditiatnai thme Hot Shutpwnmdentto ofCLMsERGAENC OLSSFICAMIONLENT L C UPGRADI~NGs ONRDN reAin EC al er townrddweh events or conditions thatcolledt meetinso excedhinghet andEA withino arlatively eshots pnterio ofsime-(ipe.,achangeading theqECLirmns I remeNT. Ifi tejdowgmedntgo the EmLi erency apprcopriateD/,th meetingL anoELdi IET then ebsdo emwergec cassificatleionshould bEAs) made ECasyfh EAasosmy bee mt.eWhileappliabl Toe aolloemrgncycasfctonlvlti approach tdonrigoremisn articularELy importammntdttedhihe emergency CLasfcto leessic ctpoidnWes aditonalitime for imLemntatio ofst protectiveomeasures d lrowngradigteELidemdaporteo thermnew teCwol then bmer bae d on a acracwihpatprocedures. Site Area Emergency with no Downgrade or terminate the emergency in long-term plant damage accordance with plant procedures. 23 Site Area Emergency with Terminate the emergency and enter recovery in long-term plant damage accordance with plant procedures. General Emergency Terminate the emergency and enter recovery in accordance with plant procedures. is pravided in P.!S 20037 02.&--2.7 CLASSIFICATION OF SHORT-LIVED EVENTS^.As di..u... in S.ectn:.2, a c Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed e.... Ifa

n o...ur. thatmet cr ........ an E. A , .......t.......mu.t.b.....lar

.radls ..f.. it.. continued presence at the time of deelo,,'a,:iein.Examples of such events include a failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip or an earthquake.CLASSIFICATION OF TRANiSIENT CONDITIONS Many of the ICs and/or EALs contained in this document employ time-based criteria.These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a.. fe ........ a fc,,... ... The following guidance should be applied to the classification of these conditions. EAL momentarily met during expected plant response -In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures. EAL momentarily met but the condition is corrected prior to an emergency declaration -If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example.An ATWS occurs and the auxiliary feedwater system fails to automatically start.Steam generator levels rapidly decrease and the plant enters an inadequate RCS heat removal condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts the auxiliary feedwater system in accordance with an EOP step and clears the inadequate RCS heat removal condition prior to an emergency declaration, then the classification should be based on the ATWS only.It is important to stress that the 15-minute emergency classification assessment period is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event;emergency classification assessments must be deliberate and timely, with no undue 24 delays. The provision discussed above addresses only those rapidly evolving situations where an operator is able to take a successful corrective action prior to the Emiergency Di-ree-tfSTED/SED completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by the plant conditions.AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITION In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1 022 is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50.72 within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements. Gui:danee on the rec-raeticn uf an. emergency deelaratien rep~ed th e NRC is dss-ezsed 25 63 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS GENERAL SITE AREA EMRENY EMRENYALERT UNUSUAL EVENT RG1 Release of RS1 Release of RA1 Release of RU1 Release of gaseous radioactivity gaseous radioactivity gaseous or liquid gaseous or liquid resulting in offsite resulting in offsite radioactivity resulting radioactivity greater dose greater than 1,000 dose greater than 100 in offsite dose greater than 2 times the (-site-mrem TEDE or 5,000 mrem TEDE or 500 than 10 mrem TEDE ...: *...mrem thyroid CDE. mrem thyroid CDE. or 50 mrem thyroid Op. Modes: All Op. Modes: All CDE., .... *Op. Modes: All limits for 60 minutes or longer.Op. Modes: All RG2 Spent fuel pool RS2 Spent fuel pool RA2 Significant RU2 UNPLANNED level cannot be level at (site-speecihl lowering of water level loss of water level restored to at least -,.esei -.patien;"l .5 above, or damage to, above irradiated fuel.(,,t" ......i Lwc";:'z 3., .. ft. (Level 3)1. irradiated fuel. Op. Modes:" Allft., Op. Modes:" All Op. Modes:" All (Level 3)1 for 60 minutes or longer.Op. Modes:" All RA3 Radiation levels that IMPEDE access to equipment necessary for normal plant operations, shutdown or cooldown.____________________________Op. Modes: All-Commented [DWS3]: V2 SFP Levels Drawing-Commented [DWS2]: V2 SFP Lneves Drawing 26 RG1 ECL: General Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE.Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)Notes:* The STED/SEDEmerga:noy Di-ectzr should declare the General Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.* The pre-calculated effluent monitor values presented in EAL # 1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. (1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Monitor Reading(WRGM rate) 2.85E+8 Time After Shutdown Reading______________ <ihr If>1hrto <2hrs RM-6481-1l* (M SL A ) 1310 mR/hr 1060 tnRihr RM-6482-1l* (MSL B) 1310 mR/hr 1060 mR/hr(MSL C) 1310 mR/hr RM-6481-2' (MSL D) 1310 mR/hr j 1060 mRlhr*With release path to the environment firom affected steam line, open ASI)V or SRV, line is faulted. or open steam supply to 1 -F W-P-37A. (site~w..; .....fi -mzn-it.cr lizt- anda .. ..... " c°)OR (2) Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE at or beyond (sitc apezifiz dcc: receptcr pzint~rhe site boundary]. OR (3) Field survey results indicate EITHER of the following at or beyond (site zpe.ific~d~r......... .pointte site boundary]: C ommentedl [DWS4]: V3 EPCALC-06-02 -Effluent Monitor Values for R EAts S- -] Commented [DWSS]: V3 EPCALC-06-02 -Efun oio' Values for R EALs 1 Commented [DWS6]: v4 0DCM and TS Basis for Site/Boundary Receptor Point S- Commented [DWS7]: V4 0DCM and TS Basis for Site[Boundary Receptor Point Closed window dose rates greater than 1000 mR/hr expected to continue for 60 minutes or longer.Analyses of field survey samples indicate thyroid CDE greater than 5000 mrem for one hour of inhalation. Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both 27 monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.Radiological effluent EALs are also included to provide a basis for classify.ing events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.28 RG2 ECL: General Emergency Initiating Condition: Spent fuel pooi level cannot be restored to at least-: ..s.t .epecific*' Leve 3.. , 1.5 ft. de....ef... jden-Lvel

3) for 60 minutes or longer.Operating Mode Applicability:

All Emergency Action Levels: INote: The Emergency DireetorSTED/SED should declare the General Emergency promptly I upon determining that 60 minutes has been exceeded, or will likely be exceeded.I (1) Spent fuel pool level cannot be restored to at least (site epecific Level 3 ;*alue~I.5 ft.above the fuel racks- for 60 minutes or longer as indicated by SF-LI-2616 (MPCS computer point A4172) or SF-LI-2617 (MPCS computer point -{ Commented [DWS8]:-V2SFP Levels Drawing Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment. lPost-Fukushima order EA-P1-051 required the installation of reliable SEP level indication capable of identifying normal level (Level I), SEP level 12 ft. 3 in. aboxe the top of the fuel racks (Level 2) and SEP level 1.5 ft. above the top of the fuel racks (Level 3).The Spent Fuel Pool Instr-umentation System (SFPIS) consists of two newv independent level instrument channels (SF-L-2616 and SF-L-26 17) in the Spent Fuel systcm. The SEPIS channels will be used to monitor spent fueI pool level during and Ibllowing beyond design basis events that could challenge the capability to ensure optimum protection for the stored fuel assemblies in the pool.Each channel is capable of measuring SEP level over a span from just above the top of the spent fuel racks to the normal SEP operating water le~el. The SFPIS will be monitored in accordanceBeyond Design Basis guidelines contained in FS~s for Extended Loss of AC' Power and Alternate SFP~ Makeup and Cooling.It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity. 29 RS1 ECL: Site Area Emergency Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE.Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)Notes:* The Emzx~rgoncy DirectorSTED/SED should declare the Site Area Emergency promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. (1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Reading RM-6528-4 (WRGM rate) 2.85E+7 uCilsec Time After Shutdown Reading<lhr >lhrto <2hrs RM-6481-1' (MSL A) 130 mR/hr 100 mRlhr RM-6482-1' (MSL B) 130 mRihr 100 mR/hr RM-6482-2' (MSL C) 130 mtlhr 100 mRihr RM-6481-2' (MSL D) 130 mR/hr 100

With release path to the environment from affected steam line, open ASDV or SRV.line is faulted, or open steam supply to I-F W-P-37A.OR (zitc zpecific monitor list and: threzho!d v'alues)(2) Dose assessment using actual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond (zitz specific ...... r ....p..... p-oinv.,the:'

site'boundary4 OR (3) Field survey results indicate EITHER of the following at or beyond (4ite-specifie-dese ... v-. v... the site boundary.Closed window dose rates greater than 100 mR/hr expected to continue for 60 minutes or longer."- Commented EDWS9]: V3 EPCALC-06-02 -Effluent Monitor/ Values for R EALs I Commented [DWSIO]: V40ODCM and TS Basis for Site[Boundary Receptor Point [DWS11]" V40ODCM and T5 Basin for Site Boundary Receptor Point Analyses of field survey samples indicate thyroid CDE greater than 500 mrem for one hour of inhalation. 30 Basis: This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via IC RG I.31 RS2 ECL: Site Area Emergency Initiating Condition: Spent fuel pool level ... ite ,,".. 'pci Lee 3 d'aezeriptic'nl.5 ft.kLevel

3) ---Commented

[DwS12]:VSF v eisr r L wviO s Operating Mode Applicability: All Emergency Action Levels: (1) Lowering of spent fuel pool level to (site: ......i ... 3 ft above the fuel racks as indicated by SF-L1-2616 (MPCS computer point A4172) or SF-LI-2617 (MPCS computer point A4220).-). -. Commented rDWSX3]: V2 SFIPLeves Draing Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IiMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Post-Fukushima order EA-12-051 required the installation of reliable SEP level indication capable of identifying normal level (Level 1). SEP level 12 ft. 3 in. above the top of the fuel racks (Level 2) and SEP level 1.5 ft. above the top of the fuel racks (Level 3).The Spent Fuel Pool Instrumentation System (SFP1S) consists of two new independent level instrument channels (SF-L-2616 and SF-I_-2617) in the Spent Fuel system. The SFPIS channels will be used to monitor spent fuel pool level during and following beyond design basis events that could challenge the capability to ensure optimum protection for the stored fuel assemblies in the pool.Each channel is capable of measuring SEP level over a span from just above the top of the spent fuel racks to the normal SEP operating water level. The SFPIS be monitored in accordance wvith Beyond Design Basis guidelines contained in FS~s for Extended Loss of AC Power and Alternate SEP Makeup and Cooling.It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity. Escalation of the emergency classification level would be via IC RG1 or RG2.32 RA1 ECL: Alert Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE.Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3 or 4)Notes:...The Emerg.n a .....torSTED/SLI) should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 15 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.* The pre-calculated effluent monitor values presented in EAL #1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available. (1) Reading on ANY of the following radiation monitors greater than the reading shown for 15 minutes or longer: Monitor Reading RM-6528-4 (WRGM rate) 2.85E+6 uCi/sec RM-6481-1" (MSL A) It) mR/hr RM-6482-1

(MSL 13) 10 mRlhr RM-6482-2" (MSL+ C) 10 mR/hr IRM-6481-2' (MSL D) It) mnRih*With release path to the environment from affe~cted steam line, open ASDV or SRV. line is faulted, or open steam supply to 1-FW-P-37A. (C:tz e-ecific mzn:^-ite

....tx dthrce^cld OR (2) Dose assessment using actual meteorology indicates doses greater than 1 0 morem TEDE or 50 mrem thyroid CDE at or beyond (site. epeci'c dccc rceptor point)~he site boundar,,j. OR (3) Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site-s~eeifie ...... e site bounday fr one hour of exposure.OR (4) Field survey results indicate EITHER of the following at or beyond 5ee recetor e~nthe site bo~undar)]:

Closed window dose rates greater than 10 mRlhr expected to continue for 60 minutes or longer.* Analyses of field survey samples indicate thyroid CDE greater than 50 mrem for one hour of inhalation.

I Comment~ed [DWS14]: V3 EPCALC-06-02 -affluent Monitor Values for R EALs SC ommntedl [DWS15]: V4 0DCM and TS Basis for Site Boundary Receptor PointCommented [DWS16]: Vs40ODCM and TS Basis for SitelBoundary Receptor Point l Commented [DWS17]: V4 00CM adT5BSi for Site SBoundary Receptor Point 33 Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a zignificant uncc~ntrclled Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE.Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Escalation of the emergency classification level would be via IC RS 1.34 RA2 ECL: Alert Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel.Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)(1) Uncovery of irradiated fuel in the REFUELING PATHWAY.OR (2) Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by high-alarm, or reading in excess of the current high-alarm setpoint on ANY of the following radiation monitors: I /M-6518-1 FSB High Range RM-6562- I FSB Vent IRM-6535A-1 Manipulator Crane RM-6535B-l1 Manipulator (3) Lowering of spent fuel pool level to (cite specific L... 'e 2 ft. 3 in. above thc fuel racks on SF-LI-261 6 (MPCS computer point A41 72) or SF-Ll-261 7 (MPCS computer point A422tI). [See n ... t ... r ....sl Basis: PATHWAY: The reactor refueling cavity, spent fuel pool and fuel transfer canal.[This IC addresses events that have caused IMMINENT or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool (see-De'el~per Noes-). These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.This IC applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with IC E-l44--EL' 1.Escalation of the emergency would be based on either Recognition Category R or C ICs.EAL # 1 This EAL escalates from RU2 in that the loss of level, in the affected portion of the REFUELING PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g.,......t fr .......n.. r ...m.... imagec), as well as significant changes in water and radiation levels, or other plant parameters. Computational aids may also be used-(e~g~., abei.!-off eiui'ie). Classification of an event using this EAL should be based on the totality of available indications, reports and observations. While an area radiation monitor could detect an increase in a dose rate due to a lowering of water 35"- Commented [DWS18]: V5 UFSAR Table 12.3-14 -CTMT tPost-LOCA Range-( Commented [DWS19]: V2 SFP Levels DrawingComamenteda [DWS20]: V6 Refueling pathway RU2 RA2 level in some portion of the REFUELING PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.EAL #2 This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel. Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event e.g., a fu..'l h....lin. ac, i ......., EAL #3 Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.Post-Fukushima order EA-12-051 required the installation of reliable SEP level indication capable of' identifying normal level (Level 1), SEP level 12 ft. 3 in. above the top of the fuel racks (Level 2) and SFP level 1.5 ft. above the top of the fuel racks (Level 3).The Spent Fuel Pool hnstrumentation System (SFP1S) consists of two new independent level instrument channels (SF-L-2616 and SF-L-2617) in the Spent Fuel system. The SFPIS channels w~ill be used to monitor spent fuel pool level during and folloxx ing beyond design basis events that could challenge the capability, to ensure optimum protection fbr the stored fuel assemblies in the pool.Each channel is capable of measuring SEP level over a span from just above the top of the spent fuel racks to the normal SEP operating xvater level. The SFPIS be monitored in accordance with Beyond Design B~asis guidelines contained in FSGs for Extended Loss of AC Po\ver and Alternate SEP Makeup and Cooling.Escalation of the emergency classification level would be via ICs RS1 or RS2-{aee--/?=2 36 RA3 ECL: Alert Initiating Condition: Radiation levels that IMPEDE access to equipment necessary for normal plant operations, shutdown or cooltdown. Operating Mode Applicability: All Emergency Action Levels: (1 or 2)INote: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. (1) Dose rate greater than 15 mRlhr in ANY of the following areas: Control Room RM6550 Central Alarm Station (CAS) by survey Secondary Alarm Station (SAS) by survey fothef OR (2) An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any of the following plant rooms or areas: related m-^de .... lizabili identified)Table H] 1)Area Mode Primary Aux Building 25 ft elevation 1 .3 7 ft elevation-26 ft elevation Turbine Building 1. 2. 3 Switchgear Rooms Essential 1, 2,3, 4 Non-essential Steam and Feedwater Pipe chases 1, 2. 3 Waste Process Building 25 ft elevation1,23 -3 ft elevation1,23 -31 ft elevation Containment

3. 4 Equipment Vaults 3. ,4

[DWS21]: V70~ps procedures 37 Basis: UNPLANNED: A parameter 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.IMPEDE; Entry into an area requires extraordinary measures to thcilitate entry of personnel into the affected room/area by installing temporary' shielding, requiring use of non-routine protective equipment. or requesting an extension in dose limits beyond normal administrative limits.This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The ..m. erg....nc... Dir......STEDiSED" should consider the cause of the increased radiation levels and determine if another IC may be applicable. For EAL #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g,, ..tal.ing tem...orar ..h......ng,reuig

u. e ... f non... r.. utin... pr..t..ti..

e-quipm...nt, requectin~g an....n....n..............d nrma dm .....trativ ........t).An emergency declaration is not warranted if any of the following conditions apply.* The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.* The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility ofaromo are (.g.,. radi..grah, ofpanroomltrrarr rez.n., ...f.r,-et,,).

The action for which room/area entry is required is of an administrative or record keeping nature (e.g., no rm',al roundes or routine inepection).
The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.38 RU1 ECL: Notification of Unusual Event Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the (ie specific ....u.nt............tr....ng de......... ODCM limits for 60 minutes or longer.Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3)Notes:* The .m.rg.n. Di... e......STED/SED should declare the Unusual Event promptly upon determining that 60 minutes has been exceeded, or will likely be exceeded.* If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded 60 minutes.* If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.(1) Reading on ANY non-isolated effluent path radiation monitor greater than 2 times the (cte .p..... effluent ...... centro.... g u' ccu .......)ODC M limits for 60 minutes or longer:(WTT Disch)RM-6521I-1 (TIB Sump)RM-6519-1 (SC B lowd own)RM-6473-l (WT LIQ kRM-6528-4 (WRGM [DWS22]: V32 UFSAR Table 11.5-1 [DWS23]: VS 012_Table 03 UFSAR WRGM/ Ranges OR (bite meniter lict nr.d thre~hcUd

.-lucez rcrr~ezndinez te 2 tim.ec the zcntrotlinr (2) Reading on ANY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.OR (3) Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the (cite epecific efflunt ....... e..n t.r........

d.. oc--.cnt)O...... V limits for 60 minutes or longer.Basis: This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (.g.,.an........le relace)... It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.39 Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.Radiological effluent EALs are also included to provide a basis for classify'ing events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions. Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.EAL # 1 -This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways.EAL #2 -This EAL addresses radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit. This EAL i4-;'l associated with planned batch releases from non-continuous release pathways EAL #3 -This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways-feg. -spi4ls-ef ............. d ino zor ,dr........,.......... ev'whanger l!elkage in river "wter eystem,z et.)....Escalation of the emergency classification level would be via IC RA I.40 RU2 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of water level above irradiated fuel.Operating Mode Applicability: All Emergency Action Levels: (1) a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following: AND b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors:, j [DWS24]: V9 SFP Level fM-6535-A-1, Containment Manipulator Crane RM-6535-B-1, Containment Manipulator Crane RM-6549-1, FSB Spent Fuel Range Lowv RM-65 18-1, FSB Spent Fuel Range HiIcommented [DWS2S]: Vio UFSAR Table 12.3-14 -CTMT I Post-LOCA Range (sue Basis: UNPLANNED: A parameter change or an event that is not l) the result of an intended e~olution or 2) an expected plant response to a transient. T'he cause of the parameter change or event may be knoxxn or unknown.REFUELING PATHWAY: The reactor refueling cavity', spent tfuel pool and fuel transfer canal.This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.A water level decrease will be primarily determined by indications from available level instrumentation. Other sources of level indications may include reports from plant personnelfrom a rcfueling crew;) or video camera observations (if available). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations. The effects of planned evolutions should be considered. For example, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading is due to an UNPLANNED loss of water level.A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refueling modes.Escalation of the emergency classification level would be via IC RA2.41 74 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS Recognition Category "C" Initiating Condition Matrix GENERAL SITE AREA EMRENY EMRENYALERT UNUSUAL EVENT CG1 Loss of reactor CS1 Loss of reactor CA1 Loss of reactor CU1 UNPLANNED vessel/RCS inventory vessel/RCS inventory vesselIRCS inventory, loss of reactor affecting fuel clad affecting core decay Op. Modes: 5, 6 vessel/RCS inventory integrity with heat removal for 15 minutes or containment capability, longer.challenged. Op. Modes: 5, 6 op. Modes: 5, 6 Op. Modes: 5, 6 CA2 Loss of all CU2 Loss of all but offsite and all onsite one AC power source AC power to to emergency buses for emergency buses for 15 minutes or longer.15 minutes or longer. Op. Modes: 5, 6, Op. Modes: 5, 6, Defueled De fueled CA3 Inability to CU3 UNPLANNED maintain the plant in increase in RCS cold shutdown, temperature. Op. Modes: 5, 6 Op. Modes: 5, 6 CU4 Loss of Vital DC power for 15 minutes or longer.Op. Modes: 5, 6 CU5 Loss of all onsite or offsite communications capabilities. Op. Modes: 5, 6, De fueled CA6 Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.______________ ______________Op. Modes: 5, 6 42 CG1 ECL: General Emergency Initiating Condition: Loss of reactor vessel/RCS inventory affecting fuel clad integrity with containment challenged. Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)SNote: The Emergency DirectcrSTED/SED should declare the General Emergency promptly I upon determining that 30 minutes has been exceeded, or will likely be exceeded.I (1) a. RVLIS Full Range < (-141.5 in) PReacter ;e'eseL/RC$ level lezs then (site....eitie -4.... ) for 30 minutes or longer.AND b. ANY indication from the Containment Challenge Table C2. (see-hele..) OR (2) a. Reactor vessel/RCS level cannot he monitored for 30 minutes or longer.AND b. Core uncovery is indicated by ANY of the following: RM-6535A-1 (Manipulator Crane) reading greater than 9500 mR/hr IRM-6535B-t (Manipulator Crane) reading greater than 9500 mR/hr(-ite-Erratic source range monitor indication UNPLANNED increase in Sumps A orPstesefi ...ee4efank... -levels of sufficient magnitude to indicate core uncovery.(the .it..... ci ............. nVisual observation. AND c. ANY indication from the Containment Challenge Table C2. ......... ,.Containment Challenge Table C2 CONTAINMENT INTEGRITY not established

Kontainment H 2 concentration

_ 6°i4(Explcsi-.e mixture) exists inside centainment UNPLANNED increase in containment pressure* If CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.Basis: UNPLANNED: A parameter 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 unknowvn.-{Commented [DWS26]: Vii EPCALC-06-04 -RVLIS Values I Commented [DWS27]: Vt0 UFSAR Table 12.3-14 Manipulator Crane Monitor Range I" Commented [DWS28]: VI3 Containment Sumnps d[ Commented [DWS29]: VI4 1-12 concentration in containment 43 CONTAINMENT INTEGRITY: The procedurally defined conditions or actions laken to secure containment and its associated structures. systems. and components as a functional barrier to fission product release under shutdown conditions. This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.With CONTAJIvMENT INTEGRITY not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT INTEGRITY is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen bumn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity. In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment, If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged. Manipulator Crane setpoint of 9500 mR/hr is 95% of the monitor range.In EAL, 2.b, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncoveiy has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. The inability to monitor reactor vesseliRCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS. RVLIS LEVEL VESSEL LEVEL (%) (inches from vessel____________ flange)-108 119.8 100 81.3 90 i 31.8 80 j -17.7 44 70 -67.2 63 -101.9 RC-LI-9405, RC-LIT-9467, 60 -116.7 and the Tygon Tube do 55 -141.5 not indicate reactor vessel 50 -166.2 level when actual level is 40 -215.7 less than -95" due to the 30 -265.2 weir on the RCP 20 -314.7 discharge. 10 -364.2 0 -413.7 _________These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal;SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industfry Actions to Assess Shutdown Management.

Reference:

FSAR Table 12.3-14 45 CS1 ECL: Site Area Emergency Initiating Condition: Loss of reactor vessel/RCS inventory affecting core decay heat removal capability. Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2 or 3)Note: The Emergency DirzztzrSTED/SED should declare the Site Area Emergency promptly upon determining that 30 minutes has been exceeded, or will likely be exceeded.(1) a. CONTAINMENT INTEGRITY not established. AND b. Reatere eLW C lve !es &"q RVLIS Full Range < 63% (- 101.9 in) 4 commented [DWS30]: VI1 EPCALC-06-04 -RVL1S Values OR (2) a. CONTAINMENT INTEGRITY established. AND b. IRVLIS Full Range < 55°/4 (-141.5{Commented [DWS31J: VII EPCALC-06-04 -RVLIS Values in u-.eactcr izevel iess tan rsite OR (3) a. Reactor vessel/RCS level cannot be monitored for 30 minutes or longer.AND b. Core uncovery is indicated by ANY of the following: RM-6535A-1 (Manipulator Crane) reading greater than 9500 m~lhr(Manipulator Crane) reading greater than 9500 mR/hr (ie Erratic source range monitor indication {4zWR}UNPLANNED increase in Kontainment Sumps A or (site-specifie zum e of sufficient magnitude to indicate core uncovery (Other zte ............. ticnVisual observation.Commented [DWS32]: V12 UFSAR Table 12.3-14 -CTMT 1 Post-LOCA Range [DWS33]: V13 Containment Sumps Basis: CONTAINMENT INTEGRITY: The procedurally defined conditions or actions taken to secure containment and its associated structures. systems, and components as a functional barrier to fission product release under shutdown conditions. UNPLANNED: A parameter 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'he known or unknown.46 This IC addresses a significant and prolonged loss of reactor vesseliRCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RCS/reactor vessel level cannot be restored, fuel damage is probable.Outage/shutdown contingency plans typically provide for re-establishing or verify'ing CONTAINMENT INTEGRITY following a loss of heat removal or RCS inventory control functions. The difference in the specified RCS/reactor vessel levels of EALs 1 .b and 2.b reflect the fact that with CONTAINMENT INTEGRITY established, there is a lower probability of a fission product release to the environment. In EAL 3.a, the 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring. Manipulator Crane setpoint of 9500 mRihr is 95% of the monitor range.The inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS. RVLIS LEVEL -VESSEL LEVEL (%) (inches from vessel____________ flange)~108 119.8 100 81.3 90 31.8 80 -17.7 70 I -67.2 63 -101.9 RC-Lt-9405, RC-LIT-9467, 60 -116.7 and the Tygon Tube do 55 -141.5 not indicate reactor vessel 50 -166.2 level when actual level is 40 -215.7 less than -95" due to the 30 -265.2 wetr on the RCP 20 -314.7 discharge. 10 -364.2 0 -413.7 _________These EALs address concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal;SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG- 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Acttons to Assess Shutdown Management. Escalation of the emergency classification level would be via IC CGI1 or RG I.47 CA1 ECL: Alert Initiating Condition: Loss of reactor vessel/RCS inventory. Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The Emergency DirectorSTED/SED should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded. I (1) Loss of reactor vessel/RCS inventory as indicated by full range < 64% (-969")vc1 .. .....il-z specifi level. l"_ - [DWS34]: V1I EPCALC-06-04 -RVL1S Values OR (2) a. Reactor vessel/RCS level cannot be monitored for 15 minutes or longer.AND b. UNPLANNED increase in ......i zump. and/cr tan4 Sumps A or levels due to a loss of reactor vesselIRCS inventory. -{ Commented [DWS35]= v13 Containment Sumps Basis: UNPLANNED: A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or exent may be known or unknow~n.This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.For EAL # 1, a lowering of water level below 640%(zite sp,-,i:ic le-el) indicates that operator actions have not been successful in restoring and maintaining reactor vessel/RCS water level.The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.Although related, EAL # 1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., lcc c, fa Recidual Heat Reea ..... ti ... * .. t..:* An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under IC CA3.For EAL #2, the inability to monitor reactor vessel/RCS level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS. 48 RVLIS LEVEL (%) 100 90 80 70 64 60 50 40 30 20 10 0 VESSEL LEVEL (inches from vessel flange)119.8" 81.3" 31.8"-17.7"-67.2"-96.9"-116.7"-166.2"-215.7"-265.2"-314.7"-364.2"-413.7" RC-LI-9405. RC-LIT-9467. and the Tygon Tube do not indicate reactor vessel level when actual level is less than -95" due to the weir on the RCP discharge. The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CSI If the reactor vessel/RCS inventory level continues to lower, then escalation to Site Area Emergency would be via IC CS1.49 CA2 ECL: Alert Initiating Condition: Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer.Operating Mode Applicability: 5, 6, Defueled Emergency Action Levels: Note:...The Emrec ... roc......ST1ED/SED should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.l For a bus to be considered energized from SEPS. both SEPS die~sel generator sets must be functional. (1) Loss of ALL offsite and ALL onsite AC Power to BOTH AC emergency buses E5 AND.... (zt .......... cme.... busec)for 1 5 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.When in the cold shutdown, refueling, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service.Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.l-EDE-SWG-5 (E5) and I-EDE-SWG-6 (E6) are the 4.16 kV emergency buses for Train A and t'rain B? respectively. These buses supply all safety-related loads.[his Initiating Condition is not met if either Bus ES or E6 is energized from the Supplemental Emergency Power System (SEPS).The SEPS primary function is to supply power to one 4.16 kV emergency bus. EDE-SWG-5 (ES) or EDE-SWG-6 (E6), in the event of a loss-of-offs ite-power (LOOP) and both [D~s fail to start and load. In addition (SEPS) provides back up power to the emergency buses when one of the emergency diesel generators (EDO) is out of service for up to fourteen days. SEPS can be used when it is anticipated that one of the ED~s will be inoperable for longer than the technical specification allowable outage time (AOT) of 72 hours.The design of the SEPS is capable of providing the required safety-related loads in the event of a loss of offsite power if both emergency diesel generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (SI. CB3S, CVI, Cl, etc.). In addition to providing power to the required loads. the total combined output of the SEPS system can supply either the RHR pump or the SI pump and one set of pressurizer heaters. These design conditions are based on Probabilistic Risk Evaluation (PRA)[[-03-007. 50 The SEPS consists of two 4.16 kV generators which use diesel fuel engines as the prime mover.The generator sets (gensets) SEPS-DG-2-A and SEPS-DG-2-B are capable of automatically starting, synchronizing together and energizing the SEPS electrical bus. The SEPS design requires a 'dead bus" transfer back to an offsite power source, i.e.. the emergency bus powered by SEPS must be dc-energized before restoring offsite power.For power restoration from the SEPS, both SEPS diesel generator sets must be functional. The use of the SEPS is recognized in the Emergency Operating Procedures.

Reference:

VESAR Section 8.3.1, AC Power Systems Escalation of the emergency classification level would be via IC C S1 or RS 1.51 CA3 ECL: Alert Initiating Condition: Inability to maintain the plant in cold shutdown.Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The Dir......STED/SED should declare the Alert promptly upon determining that the applicable time has been exceeded, or will likely he exceeded.(1) UNPLANNED increase in RCS temperature to greater ...... e-seei;ie Technical Specificat.... col t..m. ra...r. limit)" for greater than the duration specified in the following table.---] Commentedl [DWS36]: V15 Cold SD Temp Limit TS Table4-( 1 -RCS Heat-up Duration Thresholds RCS StatusCONTAINMENT INTEGRITY Ha-pDrto RCS Status ~~Status Ha-pDrtoreactor vessel >- Not applicable 60 minutes*36 inches .*W4_____________________________ Not INTACT reactor vessel Established 20 minutes*mce .........Established 0 minutes* tf an: RCS he-at zyztzma.RlI R is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

Comete [ -S3] Vt eue net OR (2) UNPLANNED RCS pressure increase greater than psig reading-). (This EAL does not apply during water-solid plant conditions.)

Basis: tUNPLANNED: A parameter change or an event that is not I) the result of an intended ev.olution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknosxn.INTACT: Capable of being pressurized. This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT INTEGRITY is established but the RCS is not intact, or RCS inventory is reduced (e.g., midwv l....p 3per-at-,n in: The 20-minute criterion was included to allow time for operator action to address the temperature increase.The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT INTEGRITY is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-52 [DWS39]: Vt7 RCS Pressure range minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety.Finally, in the case where there is an increase in RCS temperature, the RCS is not intact or is at reduced inventory [PWR], and CONTAINMENT INTEGRITY is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2)there is reduced reactor coolant inventory above the top of irradiated fuel.EAL #2 provides a pressure-based indication of RCS heat-up. The wide-range RCS pressure transmitters have a range of 0 to 3,000 psig. The main control boards have two post-accident monitoring qualified meters. one for each wide-range RCS pressure transmitter. ['hese meters have major divisions at 100 psig intervals and minor divisions at 50 psig intervals. Since it is possible to read the approximate mid-point between minor divisions, the value is set to 25 psig.Re ference: OS 1000.09 Escalation of the emergency classification level would be via IC CS 1 or RSl1.53 CA6 ECL: Alert Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode.Operating Mode Applicability: 5, 6 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: I. Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM needed for the current operating mode.OR 2. The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure needed for the current operating mode.Basis: 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.EXPLOSION: A rapid. violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (fr'om 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 detenrmine if the attributes of an explosion are present.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. SAFETY SYSTEM: A system reqtuired for safe plant operation. cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. Systems classified as safety-related.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 safety of the plant.54 EAL 1 .b. 1 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. -EAL 1 .b.2 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 would be via IC CS1 or RS1.55 CUl ECL: Notification of Unusual Event Initiating Condition: UNPLANNED loss of reactor vessel/RCS inventory for 15 minutes or longer.Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The ......nz Di... z......STIED!SED should declare the Unusual Event promptly upon I determining that 15 minutes has been exceeded, or will likely be exceeded.( I) UNPLANNED loss of reactor coolant results in reactor vessel/RCS level less than a required lower limit of an operating band, specified by an operating procedure for 15 minutes or longer.OR (4-)(2) a. Reactor vessel/RCS level cannot be monitored.-AND)b. UNPLANNED increase in (s:te ump ..... zr ........ontainment Sump A or levels.e. --(Commented [DWS40]: V13 Containment Sumnps Basis: UNPLANNED: A parameter change or an event that is not I) the result of an intended evolution or 2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.This IC addresses the inability to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a loss of the ability to monitor reactor vesselJRCS level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.Refueling evolutions that decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the core covered.EAL #1 recognizes that the minimum required reactor vessel/RCS level can change several times during the course of a refueling outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.EAL #2 addresses a condition where all means to determine reactor vessel/RCS level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against 56 other potential sources of water flow to ensure they are indicative of leakage from the reactor vessel/RCS. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CAl or CA3.57 CU2 ECL: Notification of Unusual Event Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer.Operating Mode Applicability: 5, 6, Defueled Emergency Action Levels: Notes:* The Emergency DirectorSTEID/SED should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.*For power restoration from the SEPS, both SEPS diesel generator sets must be functional. (1) a. AC power capability to Both AC emergency buses E5 ANI) E6'* oi v ......eme~enc bues)is reduced to a single power source for 15 minutes or longer.AND b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS.Basis: SAFETY SYSTEM: A system required for safe plant operation, cooling down the ptant and/or placing it in the cold shutdown condition, including the ECCS. Systems classified as safety-related.This IC describes a significant degradation of offsite and onsite AC power sources 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. When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service.Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential degradation of the level of safety of the plant.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition ~ presented below.A loss of all offsite power v,'ith a concu~ent failure of all but one emergency power source (~an oncite diesel generator). A loss of all offeite power and loss of all emergency power sources (e.g., onsite diesel P. loss 31 emergency power sources ~e.g., onsite dieset generators) wim a singte train or genemtors) with a single ~in of emergency buses being back fed from the unit main generalI I J wJ U *.1 m I *emerg..nc. buse being back;. from a.. .....t p* ower .ource.Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.58 1-EDE-SWG-5 (E5) and l-EDE-SWG-6 (E6) are the 4.16 kV emergency buses for Train A and Train B respectively. These buses supply all safety-related loads.This Initiating Condition is not met if either Bus E5 or E6 is energized from the Supplemental Emergency Power System (SEPS).The SEPS primary function is to supply power to one 4.16 kV emergency bus. EDE-SWG-5 (ES) or EDE-SWG-6 (E6), in the event of a loss-of-off'site-power (LOOP) and both EDGs fail to start and load. In addition (SEPS) provides back up power to the emergency buses when one of the emergency diesel generators (EDG) is out of service for up to fourteen days. SEPS can be used when it is anticipated that one of the EDGs will be inoperable for longer than the technical specification allowable outage time (AOl) of 72 hours.The design of the SEPS is capable of providing the required safety-related loads in the event of a loss of offsite power if both emergency diesel generators fail to start and load. During these events it is assumed that there is no seismic event or an event that requires safeguards actuation (SI, CBS. CVI, Cl, etc.). In addition to providing power to the required loads, the total combined output of the SEPS system can supply either the RHR pump or the SI pump and one set of pressurizer heaters. These design conditions are based on Probabilistic Risk Evaluation (PRA)EE-03 -007.The SEPS consists of two 4.16 kV generators which use diesel fuel engines as the prime mover.The generator sets (gensets) SEPS-DG-2-A and SEPS-DG-2-B are capable of automatically starting, synchronizing together and energizing the SEPS electrical bus. The SEPS design requires a "dead bus" transfer back to an offsite power source, i.e., the emergency bus powered by SEPS must be dc-energized before restoring offsite power.For power restoration from the SEPS. both SEPS diesel generator sets must be functional. The use of the SEPS is recognized in the Emergency Operating Procedures.

Reference:

IJFSAR Section 8.3.1, AC Power Systems The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.59 CU3 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED increase in RCS temperature. Operating Mode Applicability: 5, 6 Emergency Action Levels: (1 or 2)Note: The Emergenzy DirecterSTED!SED should declare the Unusual Event promptly upon I determining that 15 minutes has been exceeded, or will likely be exceeded. I (1) UNPLANNED increase in RC S temperature to greater than --( Commented [DWS41 ]: v15 cold SD Temp Limit TS OR (2) Loss of ALL RCS temperature and reactor vessel/RCS level indication for 15 minutes or longer.Basis: LUNPLANNEID: A parameter 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 evcnt may he knowvn or unknown.This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit, or the inability to determine RCS temperature and level, represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT INTEGRITY is not established during this event, the Emiergeshey DieeS rED/SEID should also refer to IC CA3.A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification. EAL #1 involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refueling evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.EAL #2 reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of power operation. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. 60 Escalation to Alert would be via IC CA! based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.61 CU4 ECL: Notification of Unusual Event Initiating Condition: Loss of Vital DC power for 15 minutes or longer.Operating Mode Applicability: 5, 6 Emergency Action Levels: Nt:The E..r... z Dir.. ec.....STED/SED should declare the Unusual Event promptly upon I determining that 15 minutes has been exceeded, or will likely be exceeded.I ( 1) Indicated voltage is less 1 05 Vite zpecific vzltag¢ "alue) on required Vital DC ---[ Commented [DWS42]: Vis UFSAR 8.3 2- DCV 105 limit buses associated with the Protected Vrain for 15 minutes or longer.Train A 11A and I1C Train 13 1113 and l ID Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or refueling mode.In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service. Thus, this condition is considered to be a potential degradation of the level of safety of the plant.As used in this EAL, "required" means the Vital DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Train A is out-of-service (inoperable) for scheduled outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Train B would require the declaration of an Unusual Event. A loss of Vital DC power to Train A would not warrant an emergency classification. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.Per DBD-ED-05. the DC bus voltage range within w~hich the 125 Volt DC system is considered operable is 105 volts minimum to t40 volts maximum. The vital DC Buses (Switchgear) are SWG-t1IA and I IC lor Train A and SWG-l1IB and I ID for Train B.

Reference:

UFSAR Section 8.3.2. IDC Power Systenm Procedure OS 1248.01. Loss of a Vital 125 VDC Bus Procedure VPRO ['5278. Loss of All Vital I)C Power DBD-ED-05, 125 VDC Systenm Depending upon the event, escalation of the emergency classification level would be via IC CAl1 or CA3, or an IC in Recognition Category R.62 CU5 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities. Operating Mode Applicability: 5, 6, Defueled Emergency Action Levels: (1 or 2 or 3)(I) Loss of ALL of the following onsite communication methods: In-Plant (PBX) Telephones Gai-Tronics Plant Radio System OR (2) Loss of ALL of the following ORO communications methods: Nuclear Alert System (NAS)Backup NAS All plant telephones Cellular telephones OR (3) (ste .p.. i.. lict ,of co, ....... on, .....h^,ds) Loss of ALL of communications methods: Emergency Notification System (ENS)All plant telephones FTS telephones in the TSC Cellular telephones [the following NRC 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., .... cf^ non lan, pr....... 3we equipm...nt, r.l..n ef .- cit 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 Commonwealth of Massachusetts and State of New Hampshire. 63 EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. 64}}

ML16068A131

Contents

Text

1.1 OPERATING

1.1 OPERATING

2.2 INITIATIiNC

2.1 FIsSIoN

3.1.1 General

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1.1 OPERATING

1.1 OPERATING

2.2 INITIATIiNC

2.1 FIsSIoN

3.1.1 General

Reference:

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ML16068A131

Text

1.1 OPERATING

1.1 OPERATING

2.2 INITIATIiNC

2.1 FIsSIoN

3.1.1 General

Reference:

Reference:

Reference:

Reference:

1.1 OPERATING

1.1 OPERATING

2.2 INITIATIiNC

2.1 FIsSIoN

3.1.1 General

Reference:

Reference:

Reference:

Reference:

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