NG-18-0090, Response to Request for Additional Information Regarding License Amendment Request (TSCR-166), Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. (: Difference between revisions

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{{Adams
{{Adams
| number = ML18212A231
| number = ML18212A229
| issue date = 07/26/2018
| issue date = 07/26/2018
| title = Response to Request for Additional Information Regarding License Amendment Request (TSCR-166), Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. (Pa
| title = Response to Request for Additional Information Regarding License Amendment Request (TSCR-166), Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. (Pa
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| case reference number = NG-18-0090
| case reference number = NG-18-0090
| document type = Response to Request for Additional Information (RAI)
| document type = Response to Request for Additional Information (RAI)
| page count = 125
| page count = 187
}}
}}


=Text=
=Text=
{{#Wiki_filter:NEI 99 QI (RevisioA
{{#Wiki_filter:ATTACHMENT 1 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT REQUEST TSCR-166 UPDATED REDLINE MARKUP OF NEI 99-01 REVISION 6 311 pages follow NEI 99*0 1 [Revision
: 6) }>Joye me er 2Q 12 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS 17 9 
&] , lepment of Deve
}>ffil 99 QI (RevisioR e) November 2012 Tal>le H 1: Reeognition CategoFY "H" Initiating Condition MatFix UNUSUAL EVENT HUl Confirmed SECURITY CONDITION or Op. A1edes: All HU2 Seismic e 1 ,rent greater than OBE Op. A1edes: All HUJ Hazardous e¥eflt Op. A1edes: All HU4 FIRE potentially degrading the level of safety of the plant. Op. },1edes: All ALERT HAl HOSTILE ACTION within the OWNER CO~ffROLLED AREA or airborne attack threat *within 30 minutes. Op. },fades: All HA5 Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdo 1 Nn. Op. ,\1edes: All Hf...(i Control Room evacuation resulting in transfer of
* Levels Emergency Action fer on-N Passive Reactors November 2012 
[THIS PA.GE IS LEFT BL1i\NK INTENTIONA.LLY]
NEI 99 01 [Revision
&] Nuclear Energy Institute Duane Arnold Energy Center (DAEC) Emergency Action Levels Technical Bases Document TBD,2018 November 2012 Nue.'-ear Energy Institute , I 77G I Street N W , &.*ite 4()(), Washingten D. C. (2Q2. 739.8()()0)
ACKNOWLEDGMENTS This document 1 Nas prepared by the Nuclear Energy Institute (l'J:EI) Emergency Action Level (EAL) Task Foree. NEI Chairpersen:
David Young Preparatien Team Larry Baker E>celon Nuelear/Corporate Craig Banner PSEG Nuclear/Salem and Hope Creek Nuelear Generating Stations/US,A, John Egdorf Dominion Generation/Ke 1.vaunee Power Station Jade Lewis Entergy Nuelear/Corporate C. Kelly Walker Operations Support Ser11ices , Inc. Review Team Chris Boone Southern Nuelear/Corporate John Callahan Xcel Energ)s'Corporate/USA Bill Chausse Enereon Services , Inc. Kent Crocker Progress Energy/Brunswick
}foclear Plant Don Crowl Duke Energy/Corporate Roger Freeman Constellation Energy Nuclear Group/Corporate Walt Lee TVA Nuclear/Corporate Ken Meade FENOC/Corporate Don Mathena }l"e)ctEra Energy/Corporate David Stobaugh EP Consulting , LLC }tick Turner Cal !away Plant/STARS Maureen Za1,valick Diab lo Canyon Power Plant/STARS NOTICE Neither NEI , nor any of its employees , members , supporting organizations , contractors , or consultants make any warranty , expressed or implied , or assume any legal responsibility for the accuracy or completeness of , or assume any liability for damages resulting from any use of , any information apparatus , methods , or process disclosed in this report or that such may not infringe privately 01,vned rights. N u e l-ee1r Ener gy In st itute , J 77 6 J Slreet A'. W , Sti'ite 4{)(), W*1Shingten D. C. (2()2. 739.8()()Q)
EXECUTl\fE
 
==SUMMARY==
Jl>ffil 99 0
EGL Assignment Attributes:
EGL Assignment Attributes:
3.1.3.B 207 ECL: Site Area E m ergency }ffil 99 QI (Re\1 isioR *i) }loyemeer 2012 HS7HS6 Initiating Condition:
3 .1. 3 .B 66 ECL: General Emergency N"El 99 0 I (RevisioA
Other conditions ex i s t w hi c h in th e judgment of the E mer ge nc y Dir ec tor warrant declaration of a S it e Area Emerge nc y. Operating Mode Applicability:
: 6) No 1 rember 2012 AR G1 Initiating Condition:
Al l E .. mple Emergency Action Level s: 1 Other conditions ex i st w hich in the jud gment of the Emerge nc y Dir ector indicat e t ha t events are in progress or hav e occ urr e d which invol ve act ual or lik e l y major failures of plant f un ct ion s n eede d fo r protection of the public or HOSTILE ACTI O N that r es ult s in int ent i o nal d amage or m a lici o u s acts , (1) toward si t e p erso nnel or eq uipm e nt that could l ead to th e lik e l y fai lur e of or , (2) th at prevent effect i ve access to equipme nt needed fo r the protection of the p ubli c. Any releases are not expected to result in expos ur e l eve l s w hich exceed EPA Protective Action Guideline expos ur e l evels b eyo nd the s it e boundar y. Definitions:
Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem T E DE or 5 , 000 mrem thyroid CDE. Operating Mode Applicability:
HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
All Emergency Action Levels: (1 or 2 or 3) Notes: Example Emergeney Aetien Levels: I
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individual s in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. PROJECTILE
* The Emergency Director should dec l are the General Emergencyevent promptly upon determining that the applicable time has been exceeded , or will likely be exceeded.
: An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: T hi s IC addresses unanti ci pat e d conditions not ad dr esse d explicitly e l sew h ere but that warrant d e claration of an emergency because conditions exist which are believed b y the E mergenc y Director to fall under the e mer ge nc y classification l eve l d esc ription for a Site Area E mer ge ncy. 208 NE! 99 01 (Re\'isioR 6) ~foYemaer 2012 HG1 ECL: General Emergency Initiating Condition:
* If an ongoing release is detected and the re l ease start t i me is unknown , assume that the release duration has exceeded 15 minutesthe specified time limit.
HOSTILE ACTION resulting in loss of physical control of the facility.
* If the effluent flow past an effluent monitor is known to have stopped due to actions to iso l ate the release path , then the effluent monitor reading is no l onger va l id for classification purposes.
I
* The pre-calcu l ated effluent monitor values presented in EAL .L l should_QDJy be used for emergency c l assification assessments until the resu l ts from a dose assessment using actua l meteorology are available. fL l I I Reading on ANY of the follovringTable R-1 effluent radiation monito rs greater than the rea a* h I " GE"£ 15 . t I 1ng s ownco umn or mmu es or anger: J Effluent MeniteF Glassifieatien
+lueshelds M0Rit0F Reaetor BllildiAg
\*eAtilatioA rad moAitor EKamaH ;3 1 4, ~1 6, '.7 1 8~ +llrbiAe B1JildiAg
\'eAtilatioA rad moAitor A<:::aA~aA
!12~ Gffgas 8tael , rad moAitor fK:amaA 9110~ Monitor Reactor Bu il ding ventilation r a d monitor (K aman 3/4, 5/6, 7 /8) Turb i n e Bu il ding ventil at i o n r a d monitor aJ (K a m a n 1/2) V) C1l l.!J l.lE+OO uci/cc l.4E+OO u c i/c c 67 GE I .Oe*OO llGi ,l ee l .Oeai=OO 1JGi l ee 4 .~ea1=0;3 llGi l ee 
:t-ffil 99 QI (ReYisioR e) }>foyember 2Ql2 Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5 , 000 mrem thyroid CDE at or beyond (site specific dose receptor point)the SITE BOUNDARY. [Preferred]
Field survey results indicate EITHER of the following at or beyond (site specific dose receptor point)the SITE BOUNDARY:
* Closed window dose rate s greater than 1 , 000 mR/hr expected to continue for 60_ minutes or longer.
* Analyses of field survey samp l es indicate thyroid CDE greater than 5 , 000 mrem for one hour of inhalation.
68 69 l>!El 99 0 I (ReYisioA
: 6) l>te\'emeer 2012 Definitions:
NEI 99 0 I (Re~*isioA
: 6) },/ovemeer 2012 SITE BOUNDARY:
That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the l icensee. Basis: This IC addresses a release of gaseous rad i oactivity that results in projected or actual offsite doses greater than or e qu a l to the EPA Protective Action Guides (PAGs). It in c lud es b ot h monitored and un-monitored releases.
Releases of this m agn itud e w ill require implementation of protective actio n s for the public. This IC is modified by a note that EAL RG 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actua l meteorological data and current radio l ogical conditions.
However, if Kaman monitor readings are sustained for 15 minutes or l onger and the required MIDAS dose assessments cannot be completed within this period, then the declaration can be made using Kaman readings PROVIDED the readings are not from an isolated flow path. Ra diolo g i ca l effl u e nt EALs are a l so included to provide a b asis for classifying events and co nd itions that ca nn ot be readily or appro pri ate l y classified on the b as i s of plant co nditi ons a lon e. T h e inclu s i o n of both plant condition and radiological effl u e nt EALs mor e fu ll y addresses the spec trum of possible acc id e nt eve nt s a nd conditions.
The TEDE dose i s set at the EPA PA G of 1 , 000 mrem w hil e the 5 , 000 mr em thyroid CDE was esta bli s h ed in consideration of the I :5 rat i o of the EPA PAG for TEDE and thyroid C D E. C l ass ifi cat i on based on efflue nt monitor readings ass um es th at a release path to the e n viro nm e nt i s esta bli shed. _If the effl u ent flow past an effl u e nt monitor i s known to have stoppe d du e to act i ons to isolate the re l ease path, then the effluent m o nit or r ead in g i s n o longer va lid for classification purposes.
If Kaman readings are not valid, field survey results may be utilized to assess this IC using EAL RG 1.3. Developer Notes: The effluent ICs/EALs are ineluded to provide a basis for elassifying events that eannot be readily elassified on the basis of plant eonditions alone. The inelusion of both types oflCs/EALs more fully addresses the speetrum of possible events and aeeidents.
While this IC may not be met absent challenges to multiple fission product barriers, it provides classification diversity and may be used to elassify events that would not reaeh the same EGL based on plant status or the fission product matrix alone. For many of the DB As analyzed in the Updated Final Safety Analysis Report, the diseriminator will not be the number of fission product barriers challenged , but rather the amount of radioactivity released to the environment.
The EPA PAGs are e>rpressed in terms of the sum of the effeetive dose equivalent (EDE) and the committed effective dose equiYalent (CEDE), or as the thyroid committed dose equiYalent (CDE). For the purpose of these IC/EALs , the dose quantity total effeetive dose equivalent (TEDE), as defined in 10 CFR &sect; 20, is used in lieu of " ... sum of EDE and CEDE .... ". 70 1-ffil 99 01 (Re\*isioA
: 6) NoYemaer 2012 The EPA PAG guidance provides for the use of adult thyroid dose conversion factors; however , some states haYe decided to base protective actions on child thyroid CDE. Nuclear power plant ICs/EALs need to be consistent with the protective action methodologies employed by the States within their EPZs. The thyroid CDE dose used in the IC and EALs should be adjusted as necessary to align with State protective action decision making criteria.
The "site specific monitor list and threshold values" should be determined 1.vith consideration of the follo*n<ing:
* Selection of the appropriate installed gaseous effluent monitors.
* The effluent monitor readings should correspond to a dose of 1,000 mrem TEDE or 5,000 mrem thyroid CDE at the " site specific dose receptor point" (consistent 1.vith the calculation methodology employed) for one hour of e>(posure.
* Monitor readings will be calculated using a set of assumed meteoro l ogical data or atmospheric dispersion factors; the data or factors se l ected for use should be the same as those employed to calculate the monitor readings for ICs AAl and ASl. Acceptable sources of this information include , but are not limited to, the RETS/ODCM and values used in the site's emergency dose assessment methodology.
* The calculation of monitor readings will also require use of an assumed release isotopic mix; the selected mix should be the same as that employed to calculate monitor readings for !Cs AAl and AS!. Acceptable sources of this information include , but are not limited to, the RETS/ODCM and values used in the site's emergency dose assessment methodology.
* Depending upon the methodology used to calculate the EAL values, there may be overlap of some Yalues between different ICs. Developers will need to address this overlap by adjusting these values in a manner that ensures a logical escalation in the EGL. The " site specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish between on site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan, and procedural methodology used to determine offsite doses and Protective Action Recommendations.
The yariation in selected dose receptor points means there may be some differences in the distance from the release point to the calculated dose point from site to site. Developers should research radiation monitor design documents or other information sources to ensure that 1) the EAL value being considered is 1.vithin the usable response and display range of the instrument , and 2) there are no automatic features that ma)' render the monitor reading invalid (e.g., an auto purge feature triggered at a particular indication level). It is recognized that the condition described by this IC may result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. In those cases, EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available.
For example , an El\L monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than appro>(imately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold.
Although the IC references TEDE, field survey results are generally available only as a " whole body" dose rate. For this reason , the field survey EAL specifies a "closed window" surYey reading. 71 NEI 9 9 g I (Re'\'isioA
<i) Noyember 2Ql2 . I ded in the generic EALs. . * , tern are not me u
* h
* the I time dose pr0Ject1on S) s b" lit , may not be vt'lt m Jnel i eatiens freffi 8 reah. 1,; 1 ity For those that <lo, the eapa . 1 t el 88 B',L *sing real d t ha"e t 1s capa * , Efuest to me u e ' Many lieense~so;o hni;al Sjleeifieations.
A lieensee "'"; red en a ease 1,y ease 1,asis. scope of the pant ec Its* approval ,,viii be cons 1
* fen systeffi """ '
* 1,, L tiffie <Iese pr~e,r1 . 1 <lee! in the genef!e,'
5* . erimeter monitoring system are not me u monitors may not be controlled Jnelieat1ens ffeffi 8 P h. al, ility Fer these tliat <lo , these fthe plant Teehnieal . d not have t 1s cap * . ,.,*thin the scope o A Many lieen_seese I, e leYel as plant eq*1pffient, er "'. :.. ffie~al er other faetors: n aoel maiota,neel te t e '""' eliogs "'"Y 1,e i ofl*eneeel
: 9) en,_,ron *steffl* approval w1ll l,e 8 ecifications. In add1t1_on, rea AL ing a perimeter momtormg Sy , li:ensee may "'q*est te ,nol*<I: !ffi en considered on a case by case as1s. . t Attributes*
3.1.4.C EGL Ass1gnmen n
* 72 Jl>JBI 99 g l (ReYisioA
: 6) Novemeer 2Q 12 AG2RG2 [Sec Dc 1;clepa lilotcs] ECL: General Emergency Initiating Condition:
Spent fuel pool level cannot be restored to at least 16.36 feet~ (site specific Level 3 description) for 60_-minutes or longer. Operating Mode Applicability:
All Example Emergency Action Levels: Note: The Emergency Director should declare the General Emergency event promptly upon determining that the applicable time 60 minutes has been exceeded , or will likely be exceeded.
R Q 2.1 Spent fuel pool level cannot be restored to at least 16.36 feeh (site specific Level 3 value) for 60 minutes or longer. Definitions:
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.
It is recognized that thi s IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.
DevelepeF Notes: In accordance with the di s cussion in Section 1.4 , NRG Order EA 12 051 , it is recommended that this IC and EAL be implemented when the enhanced spent fuel pool level instrumentation is available for use. The " site sp e cific L e vel 3 value" is usually that spent fuel pool level where fuel remains covered and actions to implement make up *water addition should no longer be deferred.
This site specific level is determined in accordance with : NRG Order EA 12 051 and NEI 12 02 , and applicable owner's group guidance.
Developers should modify the EAL and/or Basis section to reflect any site specific constraints or limitations associated with the design or operation of instrumentation used to determine the Level 3 &#xa5;aHie-: E GL Assignment Attributes:
3 .1.4 .G 73 NEI 99 0 I (Re\*i sioA e) November 2012 7 COLD SHUTDOWN/
REFUELING SYSTEM MALFUNCTION ICS/EALS Table C 1: Ree0gniti0n Categerv "C" Initiating C0nditi0n Matrix UNUSUAL EVENT CUl UNPLi\J,J1'ffiD loss of (reactor vesse l/RC&#xa3; [PWR] or RPV [BWR]) inventory for 15 minutes or longer. Op. 1\1edes: Geld Shutdewn , Refueli~ fr CU2 Loss of all but one AC power souree to emergency buses for 15 m i nutes or longer. Op. Afedcs: LJCeld S/n,ttde,rn, Re/Meling , Defueled CUJ ill-l"PLA}J1'1ED increase in RC&#xa3; temperature. Op. },1edcs: LJCeld Shuldew;'l , Refueling CU4 Loss of Vital DC po*n<er for 15 minutes or longer. Op. },fedcs: LJCeld Sh1itdewn , Refueling CUS Loss of all onsite or offsite communications capabilities. Op. },fedes: LJCeld 8hutdov,rn , Re.fitelir1g , Dafaeled ALERT CAl Loss of (reactor Yessel/RC8
[P WR] or R.0 V [BWR]) inventory. Op. 1\1edc s: LJCeld Shutdewn , Refiwling CA2 Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer. Op. },1edcs: LJCeld Shutdewn , Refueling , Defiwled CAJ Inability to maintain the plant in cold shutdown. Op. },1edc s: LJCeld Shutdewn , Refueling 74 8ITEAREA GENERAL EMERGENCY EMERGENCY C81 Loss of (reactor vessel/RC&#xa3;
[PWR] or fil>V [BWR]) CG 1 Loss of (reactor vessel/RC&#xa3;
[PWR] or RPV [BWR]) inventory affecting core decay heat removal capability.
Op. 1\1edcs: LJCeld Shutdewn , Refueling inventory affecting fuel elad integrity v,ith containment challenged.
Op. A1edcs: LJCeld Shi1tde*,im , Refueling
,-------------------, 1 Table intenEieEi for use by 1 I I 1 EAL Se\*elopers.
1 : Inclusion in licensee 1 I S ' , ,l 1 ocuments 1s not requ1reu. 1 L------------------*
UNUSUAL EVENT ALERT CA(i Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Op. ,\1edcs: LJCeld Shuu/ewn , Refueling 75 8ITEAREA EMERGENCY I NEI 99 QI (ReYisioR e) No,*ember 2() 12 GENERAL EMERGENCY Table intended for use b;' 1 EAL developers.
: Inclusion in licensee : documents is not required. L------------------1 NE! 99 QI (ReYisioR
: 6) 1'Jo1rember 2Q12 CU1 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED loss of (reactor vessel/RCS
[PWR] or RPV [BWR]) inventory for 15 minute s or lon g er. Operating Mode Applicability:
Cold Shutdown , Refueling4, 5 Emergency Action Levels: Example Emergeney Aetien Levels: (1 or 2) Note: The Emergency Director should declare the Unusual Eventevent promptly upon determining that the applicable time 15 minutes has been exceeded , or will lik ely be exceeded.
UNPLANNED loss of reactor coolant results .fin (reactor vessel/RCS
[PWR] or RPV [BWR]) level less than a required lower limit for 15 minutes or longer. a. (Reactor vessel/RCS
[PWR] or RPV_ [BWR]) level cannot be monitored. --AND --+---b. UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool increase in (site specific sump and/or tank) Suppression Pool or Drywell and Reactor Building floor and equipment drain sump levels. Definitions:
UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected p l ant response to a transient.
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inabi li ty to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a Joss of the ability to monitor (reactor vessel/RCS
[PWR] er RPV [BWR]) level concurrent with indications of coo lant 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 l evel decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is avai labl e to keep the core covered. 76 NEI 99 Ql (ReYisioR
: 6) 1-fo*,remser 2Q 12 EAL CUl.1 recognizes that the minimum required (reactor vessel/RCS
[PWR] or RPV [BWR]) 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. 77 78 *NEI 99 QI (ReYisioR e) ~fo\'emaer 2Q 12 
~JEJ 99 0 I (RevisioR
: 6) November 2012 ---EAL CUI .2 addresses a condition where all means to determine (reactor vessel/RCS
[PWR] or RPV [BWR]) 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 eYaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor 'vessel/RCS
[PWR] or R..1 H/ [BWR]). If all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RSC inventory loss was occurring by observing sump and Suppression Pool level changes. The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building. A rise in Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Pool level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA l or CA3. De&#xa5;elopeF Notes: EAL #1 It is recognized that the minimum allmvable reactor vessel/RCS/RP\'
level may have many values over the course of a refueling outage. Developers should solicit input from licensed operators concerning the optimum wording for this EAL statement.
In particular , determine if the generic wording is adequate to ensure accurate and timely classification , or if specific setpoints can be included without making the EAL statement umYieldy or potentially inconsistent with actions that may be taken during an outage. If specific setpoints are included , these should be drawn from applicable operating procedures or other controlling documents.
E AL #2.b E nter any " site specific sump and/or tank" levels that could be e>(pected to increase ifthere were a loss of inventory (i.e., the lost inventory would enter the listed sump or tank). EGL Assignment Attributes:
3.1.1.A 79 ECL: Notification of Unusual Event NEI 99 QI (RevisieA
: 6) ]'Jeyemeer 2012 CU2 Initiating Condition:
Loss of all but one AC power source to emergency essential buses for 15 minutes or longer. Operating Mode Applicability:
Cold Shutdown, Refuelingi,2 , Defueled Example Emergency Action Levels: Note: Th e E mergency Director should declare the Unusual Eventevent promptly upon determining that 15 minutes has been exceeded , or will likely be exceeded. c&#xb5;2.1 a. AC power capability to (site specific emergency buses) 1 A3 and l A4 buses is reduced to a sin g le power source for 15 minutes or longer. AND b. Any additional single power source failure will result in loss of all-ALL AC power to SAF E TY SYSTEMS. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition.
including the ECCS. These systems are classified as safety-related
.A system required for safe plant operation, cooling dov,rn the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems classified as safety related. Basis: This IC de s cribes a significant degradation of offsite and onsite AC power sources such that any additional s ingle failure would result in a loss of all AC power to SAFETY SYSTEMS. ln this condition , the sole AC power source may be powering one , or more than one , train of safetyrelated 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 emer g ency bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one emergency power source ( e.g., an onsite diesel generator).
* A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency essential buses being back fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a single train of 80 emergency essential buses be i ng -eae-k-fed from an offsite power source. 81 :t-1:el 99 g 1 (Re&#xa5;isi0R e) }fo>,*emaer 2Q 12 NE! 99 Q 1 (Re\*ision e) }fo*reA'!aer 2Ql2 Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. 82 NEJ 99 Q 1 (Re\1 isioA 6) 1-Jovemeer
?Q 12 Developer Notes: For a po'1.*er source that has multiple generators, the EAL and/or Basis section should reflect the minimum number of operating generators necessary for that source to provide required po'w*Jer to an AC emergency bus. For example, if a backup po'IJer source is comprised of h.vo generators (i.e., tv.*o 50% capacity generators sized to feed 1 AC emergency bus), the EAL and Basis section must specify that both generators for that source are operating.
The "site specific emergency buses" are the buses fed by offsite or emergency AC po'.ver sources that supply po'.*.*er to the electrical distribution system that po'wvers SAFETY SYSTEMS. There is typically 1 emergency bus per train of SAFETY SYSTEMS. Developers should modify the bulleted examples provided in the basis section, above, as needed to reflect their site specific plant designs and capabilities.
The EALs and Basis should reflect that each independent offsite po'w\'er circuit constitutes a single po'wver source. For example, three independent 345k'I offsite power circuits (i.e., incoming po'wver lines) comprise three separate po'1Jer sources. Independence may be determined from a review of the site specific UFSAR, 83 
~rn r 9 9 01 (R ev i s ioR 6) Nov em e er 2012 SBO analysis or related loss of electrical po'.ver studies. The EAL and/or Basis section may specify use of a non safety related po 1.ver source provided that operation of this source is recognized in AOPs and EOPS, or beyond design basis accident response guidelines (e.g., FLEX support guidelines).
Such po'.ver sources should generally meet the "Alternate ac source" definition provided in 10 CFR 50.2. At multi unit stations, the EALs may credit compensatory measures that are proceduralized and can be implemented 1.vithin 15 minutes. Consider capabilities such as power source cross ties, "sv.*ing" generators, other power sources described in abnormal or emergency operating procedures, etc. P lants that have a proceduralized capability to supply offsite AC po 1.-1er to an affected unit via a cross tie to a companion unit may credit this po'wver source in the EAL pro 1 1ided that the planned cross tie strategy meets the requirements of 10 CFR 50.63. ECL Assignment Attributes:
3.1.1.A 84 ECL: Notification of Unusual Event Initiating Condition:
UNPLANNED increase in RCS temperature.
Operating Mode Applicability:
Operating Mode Applicability:
All Example Emergency Action Level s: H 1.1 a. A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the (site specific security shift supervision)DAEC Security Shift Supervision. AND b. EITHER of the following has occurred:
Cold Shutdown, RefuelingU Emergency Action Levels: NEI 99 0 l (Re*,isioH
: 1. ANY of the following safety functions cannot be controlled or maintained.
: 6) ~foyemeer 2012 CU3 Example Emergeney Aetien Levels: ( 1 or 2) Note: The Emergency Director should declare the Unusual Eventevent promptly upon determining that the applicable time 15 minutes has been exceeded , or will likely be exceeded.  
* Reactivity control
~1 ~2 UNPLANNED increase in RCS temperature to greater than (site specific Technical Specification cold shutdown temperature limit)212&deg;F. Loss of ALL RCS temperature and (reactor vessel/RCS
* Core cooling [PWR] I RPV water level [BWR]
[PWR] or RPV [BWR]) level indication for 15 minutes or longer. Definitions:
* RCS heat removal OR 2. Damage to spent fuel has occurred or is IMMINENT.
UNPLANNED:
Definitions:
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.
HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
The cause of the parameter change or event may be known or unknown. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. For DAEC, this is considered to be Secondary Containment as required by Technical Specifications. CONTAil'J:MENT CLOSURE: 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.
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOST AGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. HOSTILE FORCE: One or more individuals who are engaged in a determined assault, overtly or by stealth and deception.
Basis: 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 CLOSURE is not established during this event , the Emergency Director 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. 85 1'ffil 99 g 1 (ReYisioR e) NoYemeer 2Q 12 EAL CU3.l 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.
equipped with suitable weapons capable of killing. maiming. or causing destruction.
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.
IMMINENT:
During an outage, the level in the reactor vessel will normally be maintained above the reactor vessel flange. Refuelin g evolutions that lower water l eve l below the reactor vessel flange are carefully planned and controlled.
The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. 209 NEI 99 QI (Re;cisioA a) November 2012 PROJECTILE:
A lo ss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.
An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. PROTECTED AREA: The area under continuous access monitoring and control, and armed protection as described in the site Security Plan. 210 Basis: 211 tffil 99 01 (Re1, 1 isioR fi) tfo1, 1 emeer 2012 NEI 99 C:l 1 (Re,*isioA 6) Novemeer 2C:ll2 This IC addresses an event in which a HOSTILE FORCE has taken physical control of the facility to the extent that the plant staff can no longer operate equipment necessary to maintain key safety functions.
86 NEI 99 0 I (ReYisioA
It also addresses a HOSTIL E ACTION leading to a loss of physical control that results in actual or IMMINENT damage to spent fuel due to 1) damage to a spent fuel pool cooling system (e.g., pumps , heat exchangers , controls , etc.) or, 2) loss of spent fuel pool integrity such that sufficient water level cannot be maintained. Timely and accurate communications between the DAEC Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan , Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].
: 6) }fo,*emeer 2012 EAL CU3.2 reflects a condition where there has been a significant loss of in strumentation capab ilit y necessary to monitor RCS conditions and operators wou ld 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 h eat l oad has been reduced since the cessation of power operation. Fifteen minutes was selected as a thresho ld to exclude transient or momentary losses of indication.
E mergency plans and implementing procedures are public documents; therefore, EALs should do not incorporate Security-sensitive information.
Escalation to A l ert would be v i a IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time cr i teria. Develof)eF Notes: for EAL #1 , enter the " site specific T e chnical Specification cold shutdovm temperature limit" where indicated.
This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
E GL Assignment Attributes:
Security-sensitive information should beis contained in non public documents such as the Security Plan. 212 Developer Notes: l>ffil 99 Q l (ReYisieA a) l>foi,*ember 2Ql 2 The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security sensitive information.
3.1.1.A 87 ECL: Notification of Unusual Event Initiating Condition:
This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location. Security sensitive information should be contained in non public documents such as the Security Plan. With due consideration given to the above developer note , EALs may contain alpha or numbered references to selected events described in the Security Plan and associated implementing procedures. Such references should not contain a recognizable description of the event. For e1rnmple , an EAL may be worded as " Security event #2, #5 or #9 is reported by the (site specific security shift supervision)." 8ee the related Developer Note in Appendix.
Loss of Vital DC power for 15 minutes or longer. Operating Mode Applicability:
B , Definitions , for guidance on the development of a scheme definition for the PROTECTED AREA. EGL Assignment Attributes:
Cold Shutdo 1 ,1rn, Refueling1_,_i Example EmergeeeyEmergency Action Levels: NEI 99 01 (ReYisioR
3 .1.4 .D 213 ECL: General Emergency l'>ffil 99 01 (ReYisioR
: 6) No,*ember 2012 CU4 Note: The E mergency Director should declare the Unusual E~vent promptly upon determining that the applicable time 15 minutes has been exceeded , or will likely be exceeded.
: 6) NoYember 2012 HG7HG6 Initiating Condition:
C 4.1 Indicated vo lt age is l ess than (site specific bus voltage value)] 05 VDC on BOTH Div 1 and Div 2 125 VDC busesrequired Vital DC buses for 15 minutes or longer. Definitions:
Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency.
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related
Operating Mode Applicability:
.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems classified as safety related. 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 availab l e 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.
All 6 E,ae,ple Emergency Action Leve ls: 1 Other conditions exist which in the judgment of the Emerge ncy Director indicate that events are in progress or have occurred which involve actua l or IMMINENT substantial core degradation or melting with potential for l oss of containment integrity or HOSTILE ACTION that results in an actua l loss of physica l control of the facility.
For example, if Train A is out-of-service (inoperable) for schedu l ed outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Tra in B would require the declaration of an Unusual Event. A loss of Vita l DC power to Train A would not warrant an emergency classification.
Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area. Definitions:
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Depending upon the event , escalation of the emergency c l assification l evel would be via IC CA 1 or CA3 , or an IC in Recognition Category AR. DeYeleper Netes: The " site specific bus voltage value" should be based on the minimum bus voltage necessary for adequate operation of SAFETY SYSTEM equipment.
HOSTILE ACTION: An act toward DAEC or its personnel that includes the use of violent force to destroy equipment, take HOSTAGES, and/or intimidate the licensee to achieve an end. This includes attack by air, land, or water using guns, explosives, PROJECTILEs, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included.
This Yoltage value should incorporate a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This Yoltage is usually near the minimum voltage selected when battery sizing is performed. 88 
HOSTILE ACTION should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on the nuclear power plant. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the owner controlled area). HOSTAGE: A person(s) held as leverage against the station to ensure that demands will be met by the station. IMMINENT:
-------~ NEI 99 01 (Re11isioR
The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions. PROJECTILE:
: 6) No11ember 2012 The typical value for aA eAtire battery set is apprmdmately 105 VDC. For a 60 cell striAg of batteries , the cell voltage is approximately 1.75 Volts per cell. For a 58 striAg battery set , the miAimum voltage is apprmdmately 1.81 Volts per cell. EGL AssigAmeAt Attributes:
An object directed toward a nuclear power plant that could cause concern for its continued operability, reliability, or personnel safety. Basis: This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a General Emergency.
3.1.1.A 8 9 NEl 99 01 (Revisien
214 11 SYSTEM MALFUNCTION ICS/EALS NEI 99 01 (ReYisieR e) }foyemaer 2012 Table S 1: Reeognition Categorv "S" Initiating Condition Matrix UNUSUAL EVENT SUl Loss of all offsite AC power capability to emergency buses for 15 minutes or longer. Op. },fades: 1. 2. 3. 4Pewer Operetien, Stertup , Het St611'lde:y, Hat 8/mtdewl'l SU2 illlPL",1'll>lED loss of Control Room indications for 15 minutes or longer. Op. ,\fades: Pewer Operetien , Stertup , Hat Stemie;*, Hat 8!1utdewnL...l,_
: 6) ~fo\cemeer 2012 ECL: Notification of Unusual Event Initiating Condition:
: 3. 4 SUJ Reactor coolant actiYity greater tkan Technical Specification allowable limits. Op. },fades: 1. 2. 3. 4Pewer Operntien, Sf6lrtup, Het Stendh;*, Het Shutdewn SU4 RCS leakage for 15 minutes or longer. Op. }.fades: 1. 2. 3, '/Pewer Operetien , Stertup, Het Stendby , Het Shutdem'l SUS Automatic or manual (trip [PWRJ / scram [BWR]) fails to shutdown the reactor. Op. }.lodes: Pewer OpaGltien}
Loss of all onsite or offsite communications capabilities. Operating Mode Applicability:
ALERT SAl Loss of all but one AC power source to emergency buses for 15 minutes or longer. Op. },fades: L...1....J...
Cold Shutdown, Refueling5, 64, 5 , Defueled Emergency Action Levels: f ... etien Le1,*els: (1 or 2 or 3) C 5.1 Loss of ALL of the following onsite communication methods: CU5 _* _(site specific list of communications methods)Plant Operations Radio System
1:.Pewer Opeffllien , Stertup , Het Sf6lndby, Het 8!1utdewn SA2 ill~PLA1'Jl>ffiD loss of Control Room indications for 15 minutes or longer witk a significant transient in progress.
Op. Mades: 1. 2. 3. 4 Pewer Operetien , Stertup , Het Stendby , Het Shu1dew1q SAS Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor, and subsequent manua l actions taken at the reactor control consoles are not successful in shutting down the reactor. Op. }.{edes: Pewer Operetienl_
2 15 SITE A.REA EMERGENCY SS1 Loss of all offsite and all onsite AC pov,rer to emergency buses for 15 minutes or longer. Op. },fades: 1, 2, 3, 4Pewer Operetien, Stertup, Het Stendb;*, Het Shutdewn SSS Inability to shutdovm the reactor causing a challenge to (core cooling [PWR] I R.0 V water level [BWR]) or RC8 heat removal. Op. }.fades: Pewer Operetienl 1 GENERAL EMERGENCY SGl Prolonged loss of all offsite and all onsite AC power to emergency buses. Op. }.fades: 1, 2. 3, 4Pewer Operetien, Stertup, Het Stendh;*, Het 8!1utdewn
-------------------
: Table intended for use by 1 EAL developers.
: Inclusion in licensee I 0 * . d , ocuments 1s not require . L------------------1 UNUSUAL EVENT SU6 Loss of all onsite or offsite eommunieations eapabilities.
Op. },fades: 1, 2, 3 , 4.Pewer Opereti e n , Stcwtup , Hat St a l'ldhy , Het Slw tde wn SU7 Failure to i s olate eontainment or loss of eontainment pressure eontrol. [PWR] Op. Afade s: 1 , 2 , 3, 1. Pewe,* Operatie,"i , Starh , tp, Het Stendhy , Hat Shi1tdewl'I A,LERT SITE AREA EMERGENCY NEI 99 01 (Re,,*isioR e) },foyemeer 2012 GENER .... L EMERGENCY SS8 Loss of all Vital DC SG8 Loss of all AC and SA9 Ha:mrdous event affeeting a SAFETY SYSTEM needed for the eurrent operating mode. Op. Med es: 1, 2 , 3, 4.Pewer Operatiel'I , Startitp , Hat S:emihy , Hat Shutdewl'1 po 1 Ner for 15 minutes or Op. },fades: 1 , 2, 3, 4Pewer Operetien , Startitp , H e t Standby*, Hat Shutdewl'I 216 Vital DC power souree s for 15 minutes or longer. Op. },1edes: 1, 2, 3, 4Pewer Operetien , St*lrt1,!fJ , Het Stendey , Hat Shutdewn ,-------------------, : Table intended for use b)' I EAL de>,<elopers.
: lnelusion in lieensee I d * . d , oeuments ts not require . 1 L------------------J ECL: Notification of Unusual Event NE! 99 0 I (ReYisioR
: 6) NoYember 2012 SU1 Initiating Condition:
Loss of alt-ALL offsite AC power capability to emergency essential buses for 15_-minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown.1..1.,_]_
Example Emergency Action Level s: Note: The E mergency Director should declare the Unusual E~vent promptly upon determining that the applicable time 15 minutes has been exceeded , or will likely be exceeded.
S 1.1 Loss of ALL offsite AC power capability to (site specific emergency buses)1A3 AND 1 A4 buses for 15 minutes or longer. Definitions:
Basis: This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency essential buses-. .,_ This condition represents a potential reduction in the level of safety of the plant. The intent of this EAL is to declare an Notification of Unusual Event when offsite power has been lost and both of the emergency diesel generators have successfully started and energized their respective 4kv essential bus. For emergency classification purposes , " capability" means that an offsite AC power source(s) is available to the emergency essential buses , whether or not the buses are powered from it. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power. E scalation of the emergency classification level would be via IC SAL De*,zeloper Notes: The " site specific emergency buses" are the buses fed by offsite or emergency AC power sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typically 1 emergency bus per train of SAFETY SYSTEMS. At multi unit stations , the EALs may credit compensatory measures that are proceduralized and can be implemented 1 within 15 minutes. Consider capabilities such as power source cross ties, " s 1.ving" generators , other power sources described in abnormal or emergenC)' operating procedures , etc. Plants that have a proceduralized capability to supply offsite AC power to an 217 NEl 99 QI (ReyisioR
: 6) tlo*,<emeer 2Q 12 affeeted unit via a eross tie to a eompanion unit may eredit this power souree in the EAL provided that the planned eross tie strategy meets the requirements of 10 CFR 50.63. EGL Assignment Attributes:
3.1.1.A 218 ECL: Notification of Unusua l Event NEI 99 QI (Re'iisioA
: 6) No&#xa5;ember 2Q 12 SU2SU3 Initiating Condition:
UNPLANNED l oss of Control Room indications for 15 minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown.L..LJ Examf)le Emergency Action Level s: Note: The Emergency Director shou ld declare the Unusual E~ve n t promptly up o n determining that the applicable time 15 minutes h as been exceeded , or w ill l ikely be exceeded.
S 3.1 ++-a.--An UNPLANNED event results in the inability to monitor one or more of the Definitions:
Reactor Power R.0 V '.Vater Level RPV Pressure Primary Containment Pressure Suppression Pool Le 1 rel
* Suppression Pool Temperature Suppression Pool Temperature Table S-1 Safety System Parameters
* Reactor power
* RPV Water Level
* RPV Pressure
* Primary Containment Pressure
* Suppression Pool Level
* Suppression Pool Temperature SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are 219 
}ffil 99 0 I (ReYisioA
: 6) N0Ye1'l'!eer 2012 classified as safety-related.A system required for safe plant operation, cooling dovm the plant and/or placing it in the cold shutdovm condition, including the EGGS. Systems classified as safety related. 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. Basis: This IC addresses the difficulty associated with monitoring normal p l ant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant. As used in this EAL , an " inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).
_For example , the reactor power level cannot be determined from any analog , dig i tal and recorder source within the Control Room. 220 NE! 99 0 I (Re,*isioH
: 6) 1-foyemeer 2012 An event involving a los s of plant indications , annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.
The event would be reported if it significantly impaired the capability to perform emergency assessments.
In particular , emergency assessments necessary to implement abnormal operating procedures , emergency operating procedures , and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.
This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control , core cooling [PWR] I RPV level [BWR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.
In addition, if all indication so urces for one or more of the listed parameters are lost , then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for reactor vessel level [PWR] / RPV water level [BWR] cannot be determined from the indications and recorders on a main control board , the SPDS or the plant computer, the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Esca lation of the emergency classification level would be via IC SA+/-}. De*,releper Netes: In the PWR parameter list column , the "site specific number" should reflect the minimum number of steam generators necessary for plant cooldown and shutdovm. This criterion may also specify whether the level value should be *wide range , narrow range or both , depending upon the monitoring requirements in emergency operating procedures.
Developers may specify either pressurizer or reactor 1 1essel level in the PWR parameter column entry for RCS Le 1 1el. The number , type , location and layout of Control Room indications , and the range of possible failure modes, can challenge the ability of an operator to accurate!)'
determine , within the time period available for emergenC)' classification assessments , if a specific percentage of indications have been lost. The approach used in this EAL facilitates prompt and accurate emergency classification assessment s by focusing on the indications for a selected subset of parameters. By focusing on the availability of the specified parameter values , instead of the sources of those values , the EAL recognizes and accommodates the wide variety of indications in nuclear power plant Control Rooms. Indication types and sources may be analog or digital , safety related or not , primary or alternate , individual meter value or computer group display , etc. /1 , loss of plant annunciators will be evaluated for reportability in accordance
*with 10 CFR 50.72 (and the associated guidance in 1'ruREG 1022), and reported if it significantly impairs the capability to perform emergency assessments.
Compensatory measures for a loss of annunciation can be readily implemented and may include increased monitoring of main control boards and more frequent plant rounds by non licensed operators.
Their alerting function notwithstanding , annunciators do not provide the parameter values or specific component status information used to operate the plant , or process through AOPs or EOPs. Based on these considerations , a loss of annunciation is considered to be adequately addressed by reportability criteria, and therefore not included in this JC and EAL. 221 
~ffil 99 QI (RevisieA
: 6) ~fo>,*emeer 2012 With respect to establishing event severity , the response to a loss of radiation monitoring data (e.g., process or effluent monitor values) is considered to be adequately bounded by the requirements of 10 CPR 50.72 (and associated guidance in NUREG 1022). The reporting of this event 1,vill ensure adequate plant staff and NRG awareness , and drive the establishment of appropriate compensatory measures and corrective actions. In addition , a loss of radiation monitoring data , by itself , is not a precursor to a more significant event. Personnel at sites that have a Failure Modes and Effects Analysis (H,4EA) included within the design basis of a digital I&C system should consider the FMEA information when developing their site specific EALs. Due to changes in the configurations of SAFETY SYSTEMS , including associated instrumentation and indications , during the cold shutdown , refueling , and defueled modes , no analogous IC is included for these modes of operation. EGL Assignment Attributes:
3.1.1.A 222 2 NEI 99 QI (ReYisioR e) Noyemeer 2Q 12 SU3SU4 ECL: Notification of Unusual Event Initiating Condition:
Reactor coolant activity greater than Technical Specification allowable limits. Operating Mode Applicability:
Power Operation , Startup, Hot Standby , Hot Shutdov,mL 2, 3 Example Emergency Action Levels: (1 or 2) (Site specific radiation monitor) reading greater than (site specific value). Pretreatment Off gas System (RM-4104)
Hi-Hi Radiation Alarm. Sample analysis indicates that reactor coolant specific activity is greater than 2.0 &#xb5;Ci/gm dose equivalent 1-131 for 12 hours or longerSample analysis indicates that a reactor coolant activity value is greater than an allowable limit specified in Technical Specifications.,,.,.
Definitions:
Basis: This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications.
This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant. For EAL SU4.1, RM-4104 Hi-Hi Radiation Alarm has been chosen because it is operationally significant is readily recognizable by the Control Room Operations Staff, and is set at a level corresponding to noble gas release rate, after 30-m i nute delay and decay of 1 Ci/sec. F or EAL SU4.2, coolant samples exceeding the 2.0 &#xb5;Ci/gm dose equivalent l-131concentration require prompt action by DAEC Technical Specifications and are representative of minor fuel cladding degradation.
Escalation of the emergency classification level would be via ICs FAl or the Recognition Category A-R I Cs. DevelepeF Netes: For EAL #1 Enter the radiation monitor(s) that may be used to readily identify 1.vhen RCS activity levels e>rneed Technical Specification allovvable limits. This EAL may be developed using different method s and sites should use existing capabilities to address it (e.g., de;, elopment of new capabilities is not required).
E>rnmples of e>C.isting methods/capabilities include:
* An installed radiation monitor on the letdown system or air ejector.
* A hand held monitor or deployed detector reading with pre calculated conversion values or readily implementable conversion calculation capability.
223 l>ffil 99 O 1 (Re*,isioR e) November 2012 The monitor reading values should eorrespond to an RCS aetivity leve l approximate l y at , Teehnieal Specification allowable limits. If there is no e>dsting method/capability for determining this EAL , then it should not be included.
IC evaluation will be based on EAL #2. For EAL#2 Developers may reword the EAL to include the reactor coolant activity parameter(s) specified in Teehnieal Specifieations and the assoeiated allowable limit(s) (e.g., va l ues for dose equ i valent I 131 and gross activity, time dependent or transient va l ues, ete.). If this approach is selected, all RCS aetivity allowable limits should be ineluded. EGL Assignment Attributes:
3 .1.1.A and 3 .1.1.B 224 NEI 99 Ql (RevisioA
: 6) November 2012 SU4SU5 ECL: Notification of Unusual Event Initiating Condition:
RCS leakage for 15 minutes or longer. Operating Mode Applicability:
Po*.ver Operation , Startup , Hot Standby , Hot Shutdownl..,_Ll Exem13le Emergency Action Levels: (1 or 2 or 3) Note: S 5.1 $2 ~3 The Emergency Director should declare the Unusual E~vent promptl y upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded.
RCS unidentified or pressure boundary leakage greater than (site specific Yalue) 10 gpm for 15 minutes or longer. RCS identified leakage greater than (site specific Yalue)25 gpm for 15 minutes or longer. Leakage from the RCS to a location outside containment greater than 25 gpm for 15 minutes or longer. Definitions:
UNISOLABLE:
An open or breached system line that cannot be isolated , remotely or locally. Basis: This IC addresses RCS leakage which may be a precursor to a more significant event. In this case , RCS leakage has been detected and operators , following applicable procedures , have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant. EAL SUS.I and EAL SU5.2 are focused on a loss of mass from the RCS due to " unidentified leakage" , "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). EAL SU5.3 addresses a RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These EALs thus apply to leakage into the containment , a secondary-side system~ steam generator tube leakage in a PWR) or a location outside of containment.
The leak rate values for each EAL were selected because they are usually observable with normal Control Room indications.
Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation).
EAL SUS.I uses a lower value that reflects the greater significance of unidentified or pressure boundary l eakage. 225 tJEI 99 Ql (RevisieR C:i) tl0 1 1em0er 2Q 12 The release of mass from the RCS due to the as-designed
/expected operation of a relief valve does not warrant an emergency classification.
For PWRs , an emergency classification would be required if a mass loss is caused by a relief valve that is not functioning as designed/expected (e.g., a relief valve sticks open and the line flo 1 t't' cannot be isolated). For BWRs , aA stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and , therefore , is not applicable to this E AL. 226 227 ~ffil 99 01 (RevisioR
: 6) ~fo*remeer 2012 NEI 99 Gl (Re,*isioA
: 6) l>fo*,*emeer 2G 12 The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage , if possible.
Escalation of the emergency classification level would be via ICs of Recognition Category A-R or F. Develof)eF Notes: Ei\L #1 For the site specific leak rate 1 1alue , enter the higher of l O gpm or the Yalue specified in the site's Technical Specifications for this type of leakage. EAL #2 For the site specific leak rate ,,atue, enter the higher of 25 gpm or the value specified in the site's Technical Specifications for this type of leakage. For sites that haYe Technical Specifications that do not specify a leakage type for steam generator tube leakage , developers should include an EAL for tube leakage greater than 25 gpm for 15 minutes or longer. EGL Assignment Attributes:
3.1. l .,", 228 
 
l>JEI 99 0 I (RevisioA
: 6) l>fo\'emeer 2012 SU5SU6 ECL: Notification of Unusual Event Initiating Condition:
Automatic or manual (trip [PWR] / scram [BWR]) fails to shutdown the reactor. Operating Mode Applicability:
Power Operationl,2 Nate: A manual action is any operator action , or set of actions , \Yhich causes the control rods to be rapidly inserted into the core , and does not include manually driving in control rods or implementation of boron injection strategies.
Exemf)le Emergency Action Levels: (1 or 2) Note: A manual action is any operator action , or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.
SU6.1 a. An automatic (trip [PWR] / scram [B'.VR]) did not shutdown the reactor. S 6.2 AND b. ANY of the following manual actions taken at 1C05 are successful in lowering reactor power below 5% power a.
* Manual Scram Pushbuttons
* Mode Switch to Shutdown
* Alternate Rod Insertion (ARI)A subsequent manual action taken at the reactor control consoles (I C05) is successful in shutting down the reactor. A manual trip ([P'.1/R]
/ scram [B'.1/R])
did not shutdown the reactor. AND b. EITHER of the following: 1. -ANY of the following subsequent manual actions taken at l COS are successful in lowering reactor power below 5% power
* Manual Scram Pushbuttons
* Mode Switch to Shutdown
* Alternate Rod Insertion (ARI)A subsequent manual action taken at the reactor control console (1 C05)s is successful in shutting down the reactor. ---__ OR 230 Definitions:
Basis: NEI 99 0 I (Re*,isioA
: 6) ~lovemeer 2012 2. -A subsequent automatic (trip [PWR] / scram [BWR]) is successful in shutting down the reactor. This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor~ [PWR] I scram [BWR]) that results in a reactor shutdown , and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [PWR] I scram [BWR]) is successful in shutting down the reactor._ This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor (trip [PWR] I scram [BWR]), operators will promptly initiate manual actions at the reactor control console s to shutdown the reactor (e.g., initiate a manual reactor (trip [PWR] I scram [BWR])). If these manual actions are successful in shutting dov,rn the reactor , core heat generation willscram quickly fall to a level within the capabilities of the plant's decay heat removal systems. 231 
---------. ------------------~JBI 99 g I (ReYisioA e) November 2012 If an initial manual reactor (trip [PWR] I scram [BWR]) is unsuccessful , operators will promptly take manual action at another location(s) on the reactor control console s to shutdown the reactor (e.g., initiate a manual reactor (trip [PWR] I scram [BWR])) using a different switch). Depending upon several factors, the initial or subsequent effort to manually (trip [PWR] I scram [BW~]) the reactor , or a concurrent plant condition , may lead to the generation of an automatic reactor ftrtp [PWR] I scram [BWR]) signal. If a subsequent manual or automatic (trip [PWR] I scram [BWR]) is successful in shutting down the reactor , core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the reactor control console s is any operator action , or set of actions , which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor ftrtp [PWR] I scram [BW1"])). This action does not include manually driving in control rods or implementation of boron injection strategies.
Actions taken at back-panels or other locations with in the Contro l Room , or any loc ation outs ide the Contro l Room , are not cons id ered to be " at the reactor control console s". Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] The plant response to the failure of an automatic or manual reactor (trip [PWR] / scram [BWR]) will vary based upon several factors including the reactor power level prior to the event , availability of the condenser , performance of mitigation equipment and actions , other concurrent plant conditions , etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutt in g down the reactor , then the emerge n cy classification l evel wi ll escalate to an Alert via IC SA:)&sect;. Depending upon the plant response , escalation is a l so possible via IC FAl. Absent the plant conditions needed to meet either IC SA:)&sect; or FAI , an Unusual Event declaration is appropriate for this event. The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power).A reactor shutdovm is determined in accordance
*.vith applicable Emergency Operating Procedure criteria.
Should a reactor (trip [PWR] / scram [BWR]) signa l be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied. I
* If the signal causes a plant transient that should have included an automatic reactor ftrtp [PWR] / scram [BWR]) and the RPS fai l s to automatically shutdown the reactor , then this IC and the EALs are applicab l e, and should be eva lu ated. If the signa l does not cause a plant trans i ent and the (trip [PWR] / scram [BWR]) failure is determined through other means (e.g., assessment oftest results), then this IC and the EALs are not applicab l e and no classification is warranted.
De*1el013er Netes: This IC is applicable in any Mode in which the actual reactor power level could eJrneed the power level at which the reactor is considered shutdown.
A PWR with a shutdown reactor power level that is less than or equal to the reactor power level which defines the lov,er bound of Pov,er Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode Applicability.
For example , if the reactor is considered to be shutdov,rn at 3% and Power Operation starts at >5%, then the IC is also applicable in Startup Mode. Developers may include site specific EOP criteria indicative of a successful reactor shutdown in 232 aA EAL staterneAt , the Basis or aoth (e.g., a reactor pov,cer leYel). wm 99 Ql (ReYisioR a) NoYemeer 2Q12 The term " reactor coAtrol coAsoles" may ae replaced with the appropriate site specific term (e.g., rnaiA eoAtrol aoards). EGL AssigArneAt Attriautes:
3.1.1.A 233 
 
1>,'EI 99 QI (Re1, 1 isioA e) l>io't'emeer 2Q 12 SU6SU7 ECL: Notification of Unusual Event Initiating Condition:
Loss of al-I-ALL onsite or offsite communications capabilities. Operating Mode Applicability:
Power Operation , Startup, Hot Standby , Hot Shutdown.L..LJ Example Emergency Action Levels: (1 or 2 or 3) S 7.1 .fill.L 2 S 7.3 Basis: Loss of ALL of the following onsite communication methods: * (site specific list of communications methods) Plant Operations Radio System
* In-Plant Phone System
* In-Plant Phone System
* Plant Paging System (Gaitronics)
* Plant Paging System (Gaitronics)
Loss of ALL of the following GRGoffsite response organization communications methods: _* _(site specific list of communications methods) DAEC All-Call phone
C l,J 5.2 Loss of ALL of the following GRGoffsite response organization communications methods:
* DAEC All-Call phone
* All telephone lines (PBX and commercial)
* All telephone lines (PBX and commercial)
* Cell Phones (including fixed cell phone system)
* Cell Phones (including fixed cell phone system)
* Control Room fixed satellite phone system
* Control Room fixed satellite phone system
* FTS Phone system Loss of ALL of the following NRC communications methods: _* _(site specific list of communications methods) FTS Phone system
* FTS Phone system (site specific list of communications methods) C 5.3 Loss of ALL of the following NRC communications methods:
* FTS Phone system
* All telephone lines (PBX and commercial)
* All telephone lines (PBX and commercial)
* Cell Phones (including fixed cell phone system)
* Cell Phones (including fixed cell phone system)
* Control Room fixed satellite phone system This IC addresses a significant loss of on-site or offsite communications capabilities.
* Control Room fixed satellite phone system * (site specific list of communications methods) 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 GRGoffsite response organization s and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible (e.g., use of non-plant , privately owned equipment , relaying of on-site information via individuals or multiple radio transmission points , individuals being sent to offsite locations , etc.). 235  
While not a direct challenge to plant or personnel safety , this event warrants prompt notifications to GRGoffsite response organization s and the NRC. 90 NEI 99 Ql (ReyisioA
 
: 6) ~loYember 2Q 12 This IC should be assessed only when extraordinary means are being utilized to make communications possible ( e.g., use of non-plant , privately owned equipment , relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations , etc.). 91  
l>IEI 99 0 I (Re*,isioA
}J:El 99 QI (RevisioA
: 6) NoYemeer 2012 EAL SU7.l addresses a total l oss of the communications methods used in support of routine plant operations. EAL SU7.2 addresses a total loss of the communications methods used to notify all GRGoffsite response organization s of an emergency declaration.
: 6) }lo,*emeer
The GRGoffsite response organization s referred to here are-the S t ate of Iowa, Linn County, and Benton County (see Developer Notes). ---EAL SU7.3 addresses a tota l loss of the communications methods used to notify the NRC of an emergency declaration.
?Q 12 EAL CU5.1 addresses a total l oss of the communications methods used in support of routine plant operations. EAL CU5.2 addresses a total loss of the communications methods used to notify a ll GRGoffsite response organization s of an emergency declaration.
DeYelopeF Notes: EAL #1 The " site specific list of communications methods" should include all communications methods used for routine plant communications (e.g., commercial or site telephones , page party systems , radios , etc.). This listing should include installed plant equipment and components , and not items owned and maintained by individuals. EAL #2 The " site specific list of communications methods" should include all communications methods used to perform initial emergency notifications to OROs as described in the site Emergency Plan. The listing should include installed plant equipment and components, and not items ovmed and maintained by individuals.
The offsite response organizations referred to here are the State of Iowa, Linn County, and Benton CountyThe OROs referred to here are (see Developer Notes). ---EAL CU5.3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.
Ei>cample methods are ring dovm/dedicated telephone lines , commercial telephone lines , radios , satellite telephones and internet based communications technology. In the Basis section , insert the site specific listing of the OROs requiring notification of an emergency declaration from the Control Room in accordance 1 Nith the site Emergency Plan , and typically within 15 minutes. EAL #3 The "site specific list of communications methods" should include all communications methods used to perform initial emergency notifications to the NRG as described in the site E,mergency Plan. The listing should include installed plant equipment and components , and not items ovmed and maintained b)' individuals. These methods are typically the dedicated Emergency Notification System (m-18) telephone line and commercial telephone lines. EGL Assignment Attributes:
Developer Notes: EAL #1 The "site specific list of coFRFRunications methods" should include all coFRFRunications FRethods used for routine plant comFRunications (e.g., coFRFRercial or site telephones, page party systeFRs, radios, etc.). This listing should include installed plant equipFRent and coFRponents, and not items owned and FRaintained by individuals.
3 .1.1.C 237 
EAL #2 The "site specific list of coFRFRunications FRethods" should include all coFRmunications FRethods used to perforFR initial emergency notifications to OROs as described in the site EFRergency Plan. The listing should include installed plant equipFRent and coFRponents, and not iteFRs owned and FRaintained by individuals.
 
farnFRple FRethods are ring down/dedicated telephone lines, coFRFRercial telephone lines, radios, satellite telephones and internet based coFRFRunications technology.
1'JE I 99 QI (R ev i s i o R e) Novembe r 2Q 1 2 SU7 ECL: Notification of Unusual Event Initiating Condition:
In the Basis section, insert the site specific listing of the OROs requiring notification of an eFRergency declaration froFR the Control RooFR in accordance with the site EFRergency Plan, and typically v,ithin 15 FRinutes.
Failure to isolate containment or loss of containment pressure control. [P\"JR] Operating Mode Applicability:
EAL #3 The "site specific I ist of comFRunications FRethods" should include all coFRmunications FRethods used to perforFR initial eFRergency notifications to the NRG as described in the site EFRergency Plan. The listing should include installed plant equipFRent and coFRponents, and not iteFRs ovmed and FRaintained by indiYiduals.
Po,.*,er Operation, Startup, Hot Standby, Hot Shutdov.*n Example Emergency Action Levels: (1 or 2) 1 a. Failure of containment to isolate 'A'hen required by an actuation signal. AND b. ALL required penetrations are not closed within 15 minutes of the actuation signal. 2 a. Containment pressure greater than (site specific pressure).
These FRethods are typically the dedicated EFRergency Notification System (ENS) telephone line and coFRFRercial telephone lines. EGL Assignment Attributes:
AND b. Less than one full train of (site specific system or equipment) is operating per design for 15 minutes or longer. Basis: Th is IC addresses a failure of one or more containment penetrations to automatically isolate (close) '.\.'hen required by an actuation signal. It also addresses an event that results in high containment pressure ,,,ith a 239 NE I 9 9 0 I (R ev i s i o R a) No,*embe r 20 1 2 concurrent failure of containment pressure control systems. Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant. For EAL 1, the containment isolation signal must be generated as the result on an off normal/accident condition (e.g., a safety injection or high containment pressure);
3.1.1.C 92 NEI 99 0 I (ReYisioR e) ~loYemaer
a failure resulting from testing or maintenance does not warrant classification.
: 20) 2 CA1 ECL: Alert Initiating Condit ion: Loss of (reactor vessel/RCS
The determination of containment and penetration status isolated or not isolated should be made in accordance v.*ith the appropriate criteria contained in the plant AOPs and EOPs. The 15 minute criterion is included to allow operators time to manually isolate the required penetrations, if possible.
[PWR] or RPV [BWR]) inventory.
EAL 2 addresses a condition
'Nhere containment pressure is greater than the setpoint at '.*Jhich containment energy (heat) removal systems are designed to automatically actuate, and less than one full train of equipment is capable of operating per design. The 15 minute criterion is included to allo'I-' operators time to manually start equipment that may not have automatically started, if possible.
The inability to start the required equipment indicates that containment heat removal/depressurization systems (e.g., containment sprays or ice condenser fans) are either lost or performing in a degraded manner. This event ,,.,ould escalate to a Site Area Emergency in accordance
,,.,ith IC FS1 if there ,.,.,ere a concurrent loss or potential loss 240 1>IBI 99 QI (Re,*isioA e) "!>lo'iemeer 2Q 12 of either the Fuel Clad or RCS fission product barriers.
Developer Notes: Enter the "site specific pressure" value that actuates containment pressure control systems (e.g., containment spray). Also enter the site specific containment pressure control system/equipment that should be operating per design if the containment pressure actuation setpoint is reached. If desired, specific condition indications such as parameter values can also be entered (e.g., a containment spray flow rate less than a certain value). EAL #2 is not applicable to the U.S. Evolutionary Po'..ver Reactor (EPR) design. Attributes:
241 ECL Assignment 3.1.1.A 242 ~m, 99 Ql (Re,*isioR e) ~fovemaer 2()12 Ne! 99 0 I (ReYisioA
: 6) :Jlolo 1 ,em0er 2012 SA1 ECL: Alert Initiating Condition:
Loss of al-l-ALL but one AC power source to emergency essential buses for 15 minutes or longer. Operating Mode Applicability:
Po 1 Ner Operation , Startup , Hot Standby , Hot Slrntdown.1.1..,_J.
Example Emergency Action Leve ls: Note: The Emergency Director should declare the Alert-event promptly upon determining that the applicable time 15_ minutes has been exceeded, or will likely be exceeded. S 1.1 a. AC power capability to (site specific emergency buses) 1A3 and 1 A4 buses is reduced to a single power source for 15 minutes or longer. AND b. Any /\NYANY additional single power source failure will result in a loss of alt ALL AC power to SAFETY SYSTEMS. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdovm condition.
including the EGGS. Systems classified as safety related. Basis: This IC describes a significant degradation of off site and onsite AC power so urces such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the so l e AC power source may be powering one, or more than one , train of safetyrelated equipment.
This IC provides an esca l ation path from IC SUL An "AC power source" is a source recognized in AOPs and EOPs , and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.
* A loss of all offsite power with a concurrent failure of all but one emergency power source ( e.g., an onsite diesel generator).
* A loss of all offsite power and loss of all emergency pO'Ner sources (e.g., onsite diesel generators) 1 Nith a single train of emergency buses being back fed from the unit main generator.
* A loss of emergency power sources (e.g., onsite diesel generators) with a sing l e train of essentialemergency buses being -eaek-fed from an offsite power source. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. Escalation of the emergency classification level would be via IC SSl. De*;eleper Notes: For a po 1 Ner source that has multiple generators , the EAL and/or Basis section should reflect the minimum number of operating generators necessary for that source to provide required power to 243 Nel 99 QI (Re\1 isioA a) l>lo*,*eme er 2Q 12 an AC emergency bus. For e>rnmple , if a backup pov,er source is comprised of two generators (i.e., two 50% capacity generators sized to feed 1 AC emergency bus), the EAL and Basis section must specify that both generators for that source are operating.
The " site specific emergency buses" are the buses fed by offsite or emergency
/\C po 1 Ner sources that supply power to the electrical distribution system that powers SAFETY 8&#xa5;8TEM8. There is typically 1 emergency bus per train of SAFETY 8&#xa5;8TEM8. Developers should modify the bulleted e>rnmples provided in the basis section , above , as needed to reflect their site specific plant designs and capabilities.
The EALs and Basis should reflect that each independent offsite pov,er circuit constitutes a single po:wer souroe. For e>rnmple, three independent 345kV offsite power circuits (i.e., incoming power lines) comprise three separate power sources. Independence may be determined from a revie\\' of the site specific UF8AR , 8BO analysis or related loss of electrical power studies. The EAL and/or Basis section may specify use of a non safety related power source provided that operation of this sou roe is recognized in AOPs and EOPs , or beyond design basis accident response guidelines (e.g., FLEX support guidelines).
8uch power sources should generally meet the " Alternate ac souroe" definition provided in 10 CFR 50.2. At multi unit stations , the E ALs may credit compensatory measures that are proceduralized and can be implemented within 15 minutes. Consider capabilities such as power source cross ties, " sv,*ing" generators , other power sources described in abnormal or emergency operating procedures , etc. Plants that have a proceduralized capability to supply offsite AC power to an affected unit Yia a cross tie to a companion unit may credit this power source in the EAL provided that the planned cross tie strategy meets the requirements of 10 CFR 50.63. EGL Assignment Attributes:
3.1.2.B 244 NEI 99 01 (Re\1 isioA e) ~fovemeer 2012 SA2SA3 ECL: Alert Initiating Condition:
UNPLANNED loss of Contro l Room indications for 15 minute s or lon ger wit h a s i gnificant transient in progress.
Operating Mode Applicability:
Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdovml.,_1,_l Example Emergency Action Level s: Note: The Emergency Director should declare the A-left-event promptly upon determining that the applicable time 15_ minutes has been exceeded , or w ill likely be exceeded.
Cold Shutdo'.&#xa5;n, Refueling4, 5 Emergency Action Levels: Example EmergeneyEmergenev Aetion LeYels: (1 or 2) Note: The Emergency Director should declare the ~event promptly upon determining that the applicable time 15_ minutes has been exceeded , or will likely be exceeded. Loss of (reactor vessel/RCS
S 3.1 a. An UNPLANNED event results in the inability to monitor one or more Table S-1 parameters from within the Control Room for 15 minutes or longerAn UNPLANl'H3D e'rent results in the inability to monitor one or more of the follo 1 Ning parameters from within the Control Room for 15 minutes or longer. ......-Table S-1 Safety System Parameters
[PWR] or RPV [BWR]) inventory as indicated by level less than (site specific level) 1 19 .5 inches. a. (Reactor vessel/RCS
* Reactor power
[PWR] or RPV [BWR]) level cannot be monitored for 15 minutes or longer AND b. UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool UNPLAN}rBD increase in (site specific sump and/or tank) levels due to a loss of (reactor vessel/RCS
* RPV Water Level
[PWR] or RPV [BWR]) inventor y. Definitions:
* RPV Pressure
UNPLANNED:
* Primary Containment Pressure
* Suppression Pool Level
* Suppression Pool Temperature PeweF RPV Water Level R1)V Pressure Primary Containment Pressure Suppression Pool Level Suppression Pool Temperature Suppression Pool Temperature AND -Reactor ___ b __ . __ ANY of the Table S-2 transient events are in progress.
245 
-Table S-2 Significant Transients
* Automatic or manual runback greater than 25% thermal reactor power
* Electrical l oad rejection greater than 25% full electrical load
* Reactor scram
* ECCS actuation
* Thermal power oscillations greater than 10% transient eyents in progress.
}>JEI 99 QI (Re\*isioA
: 6) }>Jo\*emeer 2Q 12 of the following Automatic or manual runback greater than 25% thermal reactor power Electrical load rejection greater than 25% full electrical load Reactor scram [BWR] / trip [PWR] EGGS (81) actuation Thermal power oscillations greater than 10% [BWR] 246 Definitions:  
~JEJ 99 0 I (Re\*isieA e) ~Je\1 eme er 2012 SAFETY SYSTEM: A system required for safe plant operation, coo l ing down the p l ant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling dov-m the plant and/or placing it in the cold shutdov,rn condition, including the EGGS. Systems classified as safety related. 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.
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. Basis: This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition , the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. 247 NEI 99 0 I (RevisioA
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses conditions that are precursors to a los s of the ability to adequately cool irradiated fuel (i.e., a precursor to a cha ll enge to the fuel c l ad barrier).
: 6) ])ofo&#xa5;emeer 2012 As used in this EAL , an " inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).
This cond iti on represents a potential substantial reduction in the level of plant safety. For EAL CAI .1 , a lowerin g of water level below (site specific level) 119.5 inches indicates that operator actions have not been successful in restoring and maintaining (reactor vessel/RCS
For example , the reactor power level cannot be determined from any analog , digital and recorder source within the Control Room. An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CPR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required.
[PWR] or RPV [BWR]) water l eve l. The heat-up rate of the coolant wi ll increase as the ava il ab l e water inventory is reduced. A continuing decrease in water level will lead to core uncovery.
The event would be reported if it significantly impaired the capability to perform emergency assessments.
Although related, EAL CA 1 .1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal ( e.g., loss of a Residu a l Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat remova l capability is eva luat ed under IC CA3. 93 NEI 99 Q 1 (Re't'isioR e) November 2012 For EAL CAl.2 , the inability to monitor (reactor vessel/RCS
In particular , emergency assessments necessary to implement abnormal operating procedures , emergency operating procedures , and emergency plan implementing procedures addressing emergency classification , accident assessment , or protective action decision-making.
[PWR] or RPV [BWR]) level may be caused by instrumentation and/or power failures , or water level dropping below the range of available instrumentation.
This EAL is focused on a selected subset of p l ant parameters associated with the key safety functions of reactivity control , core cooling [PWR] I RPV level [BWR] and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition.
If water level cannot be monitored , operators ma)' 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 1.vater flow to ensure they are indicative of leakage from the (reactor vessel/RCS
In addition , if all indication sources for one or more of the listed parameters are lost, then the ability to determ i ne the values of other SAFETY SYSTEM parameters may be impacted as well. For example , if the value for reactor vessel level [PW~] I RPV water level [BWR] cannot be determined from the indications and recorders on a main control board , the SPDS or the p l ant computer , the availability of other parameter values may be compromised as well. Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication. Escalation of the emergency classification level would be via I Cs FS 1 or IC AS+RS 1. Developer Notes: In the PWR parameter list column , the " site specific number" should reflect the minimum number of steam generators necessary for plant cooldown and shutdown. This criterion may also specif)' \\'Aether the level value should be vride range, narrow range or both , depending upon the monitoring requirements in emergency operating procedures. Developers may specify either pressurizer or reactor vessel level in the P'.VR parameter column entry for RCS Level. Developers should consider if the "transient events" list needs to be modified to better reflect site specific plant operating characteristics and eJ(pected responses.
[PWR] or RPV [BWR]).the operators wou l d need to determine that RSGRCS inventory loss was occurring by observing sump and Suppression Pool level changes. The drywe ll floor and equipment drain sumps, reactor building eq u ipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor bui l ding. A rise in Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Poo l l evel increases must be eva l uated against other potential sources of l eakage such as cooling water sources i nside the containment to ensure they are indicative of RCS leakage. 94 tffil 99 g I (ReYisioR e) tJoyemeer
The number, type , location and layout of Control Room indications , and the range of possible failure modes , can challenge the ability of an operator to accurately detennine , within the time period available for emergency classification assessments , if a specific percentage of indications have been lost. The approach used in this EAL facilitates prompt and accurate emergency classification assessments by focusing on the indications for a selected subset of parameters.
?QJ2 The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS 1 If the (reactor vessel/RCS
By focusing on the availability of the specified parameter values , instead of the sources of those values , the EAL recognizes and accommodates the 1.,*ide variety of indications in nuclear power plant Control Rooms. Indication types and sources may be analog or digital, safety related or not , primary or alternate , individual meter value or computer group display , etc. 248 Ne! 99 QI (RevisioR e) Jlolovemeer 2Q 12 A loss of plant annunciators will be evaluated for reportability in accordance with 10 CPR 50.72 (and the associated guidance in NUREG 1022), and reported if it significantly impairs the capability to perform emergency assessments.
[PWR] or RPV [BWR]) inventory level continues to lower , then escalation to Site Area Emergency would be via IC CS 1. Developer Notes: For EAL #1 the " site specific level" should be based on either: * [BWR] Low Low EGGS actuation setpoint/Level
Compensatory measures for a loss of annunciation can be readily implemented and may include increased monitoring of main control boards and more frequent plant rounds by non licensed operators.
: 2. This setpoint was chosen because it is a standard operationally significant setpoint at v,hich some (typically high pressure EGGS) injection systems would automatically start and is a value significantly below the low RPV v,ater level RPS actuation setpoint specified in IC GU 1. * [PWR] The minimum allowable level that supports operation of normally used decay heat removal systems (e.g., Residual Heat Removal or Shutdown Cooling).
Their alerting function notwithstanding , annunciators do not provide the parameter values or specific component status information used to operate the plant , or process through AOPs or EOPs. Based on these considerations , a loss of annunciation is considered to be adequately addressed by reportability criteria , and therefore not included in this IC and EAL. With respect to establishing event severity , the response to a loss of radiation monitoring data (e.g., process or effluent monitor values) is considered to be adequately bounded by the requirements of 10 CPR 50.72 (and associated guidance in NUREG 1022). The reporting of this event will ensure adequate plant staff and NRG awareness , and drive the establishment of appropriate compensatory measures and corrective actions. In addition , a loss of radiation monitoring data , by itself , is not a precursor to a more significant event. Personnel at sites that have a Failure Modes and Effects Analysis (FMEA) included within the design basis of a digital J&C s;'stem should consider the FMEA information when developing their site specific EALs. Due to changes in the configurations of 8AFETY 8&#xa5;8TEM8 , including associated instrumentation and indications , during the cold shutdown , refue l ing , and defueled modes , no analogous IC is included for these modes of operation.
If multiple levels e>,ist , specify each along with the appropriate mode or configuration dependency criteria.
For EAL #2 The type and range of RCS level instrumentation may vary during an outage as the plant moves through various operating modes and refueling evolutions, particularly for a PWR. As appropriate to the plant design, alternate means of determining RCS level are installed to assure that the ability to monitor level v,ithin the range required by operating procedures will not be interrupted.
The instrumentation range necessary to support implementation of operating procedures in the Gold Shutdovm and Refueling modes may be different (e.g., narrower) than that required during modes higher than Gold Shutdown.
E nter any " site specific sump and/or tank" levels that could be expected to increase if there were a loss of inventory (i.e., the lost inventOF)' would enter the listed sump or tank). EGL Assignment Attributes:
3.1.2.B 95 ECL: Alert 1'ffil 99 QI (Re,*isioR 6) 1'IO'ieFReer 2Q 12 CA2 Initiating Condition:
Loss of all offsite and all onsite AC power to emergency essential buses for 15 minutes or longer. Operating Mode Applicability:
Cold Shutdown , Refueling4.
5 , Defueled Emergency Act i on Levels: Example EmergeneyEmergeney A.etien Le11els: Note: The Emergency Director should declare the A-left-event promptly upon determining that the applicable time 15_ minutes has been exceeded , or will likely be exceeded. C 2.1 Loss of ALL offsite and ALL onsite AC Power to (site specific emergency buses)1A3 and 1 A4 buses for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A system required for safe plant operation.
cooling down the plant and/or placing it in the co l d shutdown condition, including the ECCS. These systems are class i fied as safety-related.A system required for safe p l ant operat i on, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. Systems classified as safety related. 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 thresho l d to exclude transient or momentary power losses. ---Escalation of the emergency classification level would be via IC CSl or A8+RSl. Develeper Netes: For a povt1er source that has multiple generators , the EAL and/or Basis section shou l d reflect the minimum number of operating generators necessary for that source to provide adequate power to an AC emergency bus. For e>rnmple , if a backup pm.ver source is comprised of two generators 96 NEI 99 g 1 (RevisioR
: 6) tlo't'emeer 2912 (i.e., two 50% eapaeity generators sized to feed 1 AC emergeney bus), the EAL and Basis seetion must speeify that both generators for that souree are operating.
The "site specific emergency buses" are the buses fed by offsite or emergency AC pov,er sources that supply power to the eleetrieal distribution system that powers SAFETY SYSTEMS. There is typieally 1 emergency bus per train of SAFETY SYSTEMS. The EAL and/or Basis section may specify use of a non safety related power source provided that operation of this souree is eontrolled in aeeordanee with abnormal or emergeney operating proeedures , or beyond design basis aeeident response guidelines (e.g., FLEX support guidelines). Sueh power sourees should generally meet the " Alternate ae souree" definition provided in 10 CFR 50.2. At multi unit stations , the EALs may credit compensatory measures that are proeeduralized and ean be implemented 1tvithin 15 minutes. Consider eapabilities such as power source cross ties, "s 1 i11ing" generators , other power sources described in abnormal or emergency operating procedures , ete. Plants that have a proeeduralized eapability to supply offsite AC power to an affected unit via a cross tie to a companion unit may credit this power source in the E AL provided that the planned eross tie strategy meets the requirements of 10 CFR 50.63. EGL Assignment Attributes:
3.1.2.B 97 NE! 99 0 I (RevisioA
: 6) }lo\*emeer 2012 CA3 ECL: Alert Initiating Condition:
Inability to m ai nt a in the plant in cold shutdow n. Operating Mode Applicability:
Cold Shutdovm , Refuel in~ Emergency Action Levels: Example Emergeney Aetien Levels: (1 or 2) Note: The E mergenc y Director s hould d e clare th e A-left-event promptly up o n determinin g that the app li ca ble time h as b ee n excee d e d , or w ill lik e l y b e excee d ed. C 3.1 C 3.2 UNPLANNE D increase in R CS temperature to greater than (site specific Technical Specification cold shutdown temperature limit)212&deg;F for grea t e r than the duration spec ifi e d in the following tableTable C-2. Table C-2+ RCS Heat-up Duration Thresholds RCS 8tatuslntegri!Y CONTAINMENT CLOSURE Heat-up Duration Status Intact (but not at reduced Not applicable 60 minutes* in*rentory
[PWR]) Not intact (or at reduced Estab li s hed 20 minute s* in*rentor)
' [PWR]) Not Esta blished 0 minute s
* _If a n RCS h eat removal syste m i s in operation within this time frame and RC S temperature i s being r e duc ed, the EAL i s not applicable.
UNPLANNED RCS pressur e increa se g reater than (site specific pressure reading) IO psig due to a loss of RCS cooling .. (This EAL does not apply during water solid plant conditions.  
[PWR]) Definitions:
UNPLANN E D: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant respon s e to a transient.
The cause of the parameter change or event may be known or unknown. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures.
systems, and components as a functional barrier to fission product release under existing plant conditions.
For DA E C, this is considered to be Secondary Containment as required by Technical Specifications.
CONTAINMENT CLOSURE: The proc e durally 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.
98 Basis: NEI 99 0 I (Re*,isioR
: 6) ~JOY8ffl08F 2012 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.
RCS integrity is intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). 99 NEI 99 QI (Re&#xa5;isioA
: 6) ~lo\*emeer 2Q 12 The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact , or RCS inventory is reduced (e.g., mid loop operation in P'.1/Rs) . .! The 20-min ut e criterion was included to allow time for operator action to address the temperature increase.
The RCS Heat-up Duration Thresho ld s table a l so addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not cruc i a l in this cond iti on since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety. Fina ll y, in the case where there is an increase in RCS temperature, the RCS is not int act or is at reduced inventory
[PWR], and CONTAINMENT CLOSURE is not estab li shed , 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 coo lant inv entory above the top of irradiated fuel. EAL CA3.2 provides a pressure-based indication of RCS heat-up. Esca l ation of the emergency c l assification l evel wou ld be via IC CS 1 or AS+RS 1. DeYeloper Notes: For EAL #1 Enter the "site specific Technical Specification cold shutdovm temperature limit" where indicated.
The RC8 should be considered intact or not intact in accordance v,ith site specific criteria.
For EAL #2 The "site specific pressure reading" should be the lowest change in pressure that can be accurately determined using installed instrumentation , but not less than 10 psig. For PWRs, this IC and its associated EALs address the concerns raised by Generic Letter 88 17, Less ofDee*ly Heat Reme-vel.
/\ number of phenomena such as pressuri2:ation, vorte,..ing, steam generator U tube draining , RCS level differences when operating at a mid loop condition, decay heat removal system design , and level instrumentation problems can lead to conditions where decay heat removal is lost and core uncovery can occur. NRG analyses shov,r that there are sequences that can cause core uncovery in 15 to 20 minutes , and severe core damage within an hour after decay heat removal is lost. The allo1,ved time frames are consistent with the guidance provided by Generic Letter 88 17 and believed to be conservative gi 1 1en that a low pressure Containment barrier to fission product release is established.
EGL Assignment Attributes:
EGL Assignment Attributes:
3.1.2.B 249 
3.1.2.B 100 ECL: Alert ~ffil 99 01 (RevisioR e) November 2012 CA6 Initiating Condition:
 
Hazardo u s event affect in g a SAFETY SYSTEM n eede d for the current o p erat in g mode. Operating Mode Applicability:
ECL: Alert }ffil 99 0 I (ReYisioA
Cold Shutdown, Refuelin~
: 6) }lo\1 emeer 2012 SA5SA6 Initiating Condition:
Emergency Action Levels: ExamJJle EmeFgeney Aetien Levels: l. Notes:
Automatic or manual (trip [P'.l/R] / scram [BWR]) fails to shutdown the reactor , and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor. Operating Mode Applicability:
* If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred.
Pov,er Operation.1.....1 Note: A manual action is any operator action , or set of actions , which causes the control rods to be rapidly inserted into the core , and does not include manually driving in control rods or implementation of boron injection strategies. l E,ample EmeFgeeeyEmergency Action Level s: 1 a. An automatic or manual (trip [PWR] / scram [BWR]) did not shutdown the reactor. AND b. ALL of the following manual actions taken at 1 COS are not successful in lowering reactor power below 5% power
then this emergency classification is not warranted.  
* Manual Scram Pushbuttons
* -If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM. then this emergency classification is not warranted.
* Mode Switch to Shutdown
C 6.1 a. The occurrence of ANY of the Table C-3 hazardous events:The occurrence of ,'\NY of the following hazardous events:
* Alternate Rod Insertion (ARI)Manual actions taken at the reactor control consoles (1 COS) are not successful in shutting down the reactor. Definitions:
* Seismic event (earthquake)
Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor ftrtJ3 [PWR] I s cram [BWR]) that results in a reactor s hutdown , and subsequent operator manual actions taken at the reactor control consoles to shutdown the reactor are also unsuccessful.
This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emer g ency declaration is required even if the reactor is s ubsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS. A manual action at the reactor control consoles is any operator action , or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor ftrtJ3 [PWR] I scram [BWR])). This action does not include manually driving in control rods or implementation of boron injection strategies.
If this action(s) is unsuccessful , operators would immediately pursue additional manual actions at locations away from the reactor control consoles (e.g., locally opening breakers).
_Actions taken at back-panels or other locations within the Control Room , or any location outside the Control Room , are not considered to be " at the reactor control consoles.,_''-:-251 NEI 99 01 (RevisioA a) l>fovemeer 2012 Taking the Reactor Mode Switch to SHUTDOWN is considered to be a manual scram action. [BWR] 252 
 
NEI 99 0 I (RevisioA
: 6) ]I,[ 0\'emeer 2012 The plant response to the failure of an automatic or manual reactor (trip [PWR] I scram [BWR]) will vary based upon several factors including the reactor power level prior to the event , availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions , etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling [PWR] / RPV water level [BWR] or RCS heat removal safety functions , the emergency classification level will escalate to a Site Area E mergency via IC SS~&sect;. Depending upon plant responses and symptoms, escalation is also possible via IC FSl. Absent the plant conditions needed to meet either IC SS~&sect; or FSl, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however , this IC and EAL are included to ensure a timely emergency declaration. The reactor should be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power).A reactor shutdovm is determined in accordance
*.vith applicable Emergency Operating Procedure criteria.
Develeper Netes: This IC is applicable in any Mode in which the actual reactor power level could exceed the pov,rer level at which t he reactor is considered shutdown.
A PWR *.vith a shutdown reactor power level that is less than or equal to the reactor power level which defines the lower bound of Pmver Operation (Mode l) will need to include Startup (Mode 2) in the Operating Mode Applicability. For example , if the reactor is considered to b e shutdovm at 3% and Power Operation starts at >5%, then the IC is also applicable in Startup Mode. Developers may include site specific EOP criteria indicative of a successful reactor shutdown in an EAL statement , the Basis or both (e.g., a reactor power level). The term " reactor control consoles" may be replaced v r ith the appropriate site specific term (e.g., main control boards). EGL Assignment Attributes
: 3.1.2.B 254 
 
Nm 99 QI (ReYisioR
: 6) ~loYemller 2Q 12 SA9SA8 E CL: A l ert Initiating Condition:
Hazar d o u s eve nt affec tin g a SAFETY SYSTEM n ee d e d for th e c u rre nt o p e ratin g m o d e. Operatin g Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdovm.1..1.,_]_
Example Emergenc y A ction Le v el s: Notes: S 8.1
* If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardot1s event occurred, then this emergency classification is not warranted. -If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of the SAFETY SYSTEM, then this emergency classification is not warranted .
* a. AND b. The occurrence of ANY of the Table S-3 hazardous events: ----l Table S-3 Hazardous Events -* Seismic event (earthquake)
* Internal or external flooding event
* Internal or external flooding event
* High winds o r t ornado strike
* High winds or tornado strike
* FIRE
* FIRE
* EXPLOSION
* EXPLOSION
* Othe r events w i th s i milar hazard c h aracteristics as determined by the Sh i ft Manager or Emergency Director Director 1. 2. Event damage has caused indications of degraded performance in one train of a SAFETY SYSTEM needed for the current operating mode. AND EITHER of the following:
* Other events with similar hazard characteristics as determined by the Shift Manager or Emergency Director
256 
~IEJ 99 Q I (Re,,isioA
: 6) *Novemeer 2Q 1 2
* Event damage has caused indications of degraded performance to a second train of t h e SAFETY SYSTEM needed for the current operating mode,
* The event has resulted in VfSTBLE DAMAGE to the second train of a SAFETY SYSTEM needed for the current operating mode.;: E 1 ,rent damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode, or The eyent has resu l ted in VISIBLE DAMAGE to the second train of the SAFETY SYSTEM needed for the current operating mode. Loss of the safety function of a single train SAFETY SYSTEM. 257 
*------------------------Definitions:
NEI 99 0 I (ReYisioA e) 1-Jo,,ember 20 12 EXPLOSION:
A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. 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.
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, inc l uding the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdov,rn condition, including the BCCS. Systems classified as safety related. VISIBLE DAMAGE: Damage to a component or structure that is readily observable without measurements, testing, or analysis.
The visua l impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected component or structure.
Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. EITHER of the following:
: 1. event damage has caused indications of degraded performance in at least one train of a SAFBTY 8Y8TBM needed for the current operating mode. OR 2. The e1,*ent has caused V18IBLB DAMAGE to a SAFBTY 8&#xa5;8TBM component or structure needed for the current operating mode. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria SA98.1.b.1 of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. An event affecting a single-train SAFETY SYSTEM (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under SA8 because the two-train impact criteria that underlie the EALs and Bases would not be met. If 2 58 Ne! 99 01 (Re;1 isioR e) ~J o*temeer 2012 a n event affects a single-train SAFETY SYST E M, then the emergency cl ass ification should be made based on plant parameters
/symptoms meeting the E ALs for another TC. Depending upon the circ u mstances, classificatio n may also occur based o n Shift Manager/Emergency Director judgement.
Ind i cations of degraded performance addresses damage to a SAFETY SYSTEM tra i n that is i n serv i ce/operation s i nce indications for it wi ll be readily available.
The indications of degraded performance should be significa n t enough to cause concern regarding the operabi l ity or reliability of the SAF E TY SYSTEM train. 259 
 
~JEI 99 QI (ReYisioA e) No,*ember 2Q 12 VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. 261 Tl-IE! 99 0 I (Re,,*isioA
: 6) November 2012 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 EAL l.b.l addresses damage to a SAFETY SYSTEM train that is in serYice/operation since indications for it 1,vill 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 eYent 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 FS 1 or Ml-RS 1. Develeper Netes: For (site specific hazards), developers should consider including other significant, site specific hazards to the bulleted list contained in EAL 1.a (e.g., a seiche). *Nuclear power plant SAFETY SYSTEMS are comprised of t>.&#xa5;0 or more separate and redundant trains of equipment in accordance with site specific design criteria.
EGL Assignment ,r\ttributes:
3.1.2.B 262 ECL: Site Area Emergency NEI 99 0 I (Re*,isioR
: 6) No*,emeer 2012 551 Initiating Condition:
Loss of ALLa-lt offsite and al-I-ALL onsite AC power to emergency essential-buses for 15 minutes or longer. Operating Mode Applicability:
Pov,*er Operation , Startup , Hot Standby , Hot Shutdownl....1.J Example Emergency Action Level s: Note: The Emergency Director should declare the Site Area Emergencyevent promptly upon determining that the applicable time 15 minutes has been exceeded , or will likely be exceeded.
S 1.1 Loss of ALL offsite and ALL onsite AC power to (site specific emergenC)' buses) 1 A3 and 1A4 buses for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A s ystem required for safe plant operation.
cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling do 1.vn the plant and/or placing it in the cold shutdovm condition, including the EGGS. Systems cla s sified a s safety related. Basis: This IC addresses a total loss of AC power that compromises the performance of a ll SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling , containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Esca l ation of the emergency c l assification l eve l wou ld be via I Cs AG+RG 1 , FG I or SG I. De*;eloper Notes: For a power source that has multiple generators , the EAL and/or Basis section should reflect the minimum number of operating generators necessary for that source to provide adequate power to an AC emergency bus. For e>rnmple , if a backup power source is comprised of two generators (i.e., tv,o 50% capacity generators sized to feed 1 AC emergency bus), the E AL and Basis section must specify that both generators for that source are operating.
The " site specific emergency buses" are the buses fed by offsite or emergency AC power sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typically l emergenC)' bus per train of SAFETY SYSTEMS. 263 _J NBT 99 0 I (RevisioA
: 6) November 2012 The EAL and/or Basis section rnay specify use of a non safety related power source pro 1 1ided that operation of this source is controlled in accordance with abnorrnal or ernergency operating procedures , or beyond design basis accident response guidelines (e.g., FLEX support guidelines).
Such power sources should generally rneet the "Alternate ac source" definition provided in 10 GFR 50.2. At rnulti unit stations , the EA.Ls rnay credit compensatory rneasures that are proceduralized and can be irnplernented within 15 minutes. Consider capabilities such as power source cross ties, " swing" generators , other po 1.Yer sources described in abnormal or ernergency operating procedures , etc. Plants that have a proceduralized capability to supply offsite AG power to an affected unit via a cross tie to a cornpanion unit may credit this po 1.ver source in the EAL provided that the planned cross tie strategy meets the requirernents of 10 GFR 50.63. EGL Assignrnent Attributes:
3.1.3.B 264 ECL: Site Area Emergency
~JEI 99 0 I (RevisioA
: 6) November 2012 SS8SS2 Initiating Condition:
Loss of ALL Vital DC power for 15 minutes or longer. Operating Mode Applicability:
l, 2, 3 Emergency Action Levels: Note: The Emergency Director should declare the Site Area Emergenc)'event promptly upon determining that the applicable time 15 minutes has been exceeded, or will likely be exceeded.
~S+=-2~. l __ Indicated voltage is less than (site specific bus voltage value) 105 VDC on ALL(site specific Vital DC busses) BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer. Definitions:
SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling do,Yn the plant and/or placing it in the cold shutdown condition, including the EGGS. S)'stems classified as safet)' related. Basis: This IC addresses a loss of Vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown , this condition involves a major failure of plant functions needed for the protection of the public. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Escalation of the emergency classification level would be via I Cs AG-1-RG 1 , FG 1 or SG2. 265 
~ml 99 Ql (RtwisioA a) No*,emaer 2Q12 SS5SS6 ECL: Site Area Emergency Initiating Condition:
Inability to shutdown the reactor causing a challenge to (core cooling [PWR] I RPV water level [BWR]) or RCS heat removal. Operating Mode Applicability:
Power OperationLl Examf)le Emergency Action Levels: S 6.1 a. b. An automatic or manual (trip [PWR] / scram [BV/R]) did not shutdown the reactor. AND ALL of the following manual actions taken at 1 COS are not successful in lowering reactor power be l ow 5% power:
* Manual Scram Pushbuttons
* Mode Switch to Shutdown
* Alternate Rod Insertion (ARI)AII manual actions to shutdown the reactor have been unsuccessful.
AND c. EITHER of the following conditions exist: Definitions:
Basis: _* _(Site specific indication of an inability to adequately remove heat from the core) Reactor vessel 'NaterRPV level cannot be restored and maintained above -25 inches. * (Site specific indication of an inability to adequately remove heat from the -RGSjHCL (Graph 4 of EOP 2) exceeded.
This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor ftFtp [PWR] I scram [BWR]) that results in a reactor shutdown , all subsequent operator actions to manually shutdown the reactor are unsuccessful , and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.
In some instances , the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition 266 NE! 99 QI (ReYisieR
: 6) "t-Je&#xa5;ember 2Q 12 Category F ICs/EALs.
This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency in response to prolonged failure to shutdown the reactor. The reactor shou l d be considered shutdown when it is producing less heat than the maximum decay heat load for which the SAFETY SYSTEMS are designed (typically 3 to 5% power). A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria. Esca lati on of the emergency c l assification le ve l would be via IC AG+-RG 1 or FG 1. De*,relof)er Notes: This IC is applicable in any Mode in which the actual reactor povt'er level could e,rneed the power level at vt'hich the reactor is considered shutdown.
A PWR with a shutdown reactor power level that is less than or equal to the reactor power level which defines the lower bound of Power Operation (Mode 1) will need to include Startup (Mode 2) in the Operating Mode Applicability.
For e,rnmple , if the reactor is considered to be shutdown at 3% and Power Operation starts at >5%, then the IC is also applicable in Startup Mode. Developers may include site specific EOP criteria indicative of a successful reactor shutdown in an EAL statement, the Basis or both (e.g., a reactor power level). Site specific indication of an inability to adequately remove h e at from the core: [BWR] Reactor vessel water level cannot be restored and maintained above Minimum Steam Cooling RPV Water Level (as described in the EOP bases). [PWR] Insert site specific values for an incore/core e,cit thermocouple temperature and/or reactor vessel water level that drives entry into a core cooling restoration procedure (or otherwise requires implementation of prompt restoration actions).
Alternately , a site may use incore/core e J lit thermocouple temperatures greater than l , 200 6 F and/or a reactor vessel water level that corresponds to apprmcimatel)
' the middle of active fuel. Plants vt'ith reactor vessel level instrumentation that cannot measure down to approximately the middle of active fuel should use the lowest on scale reading that is not above the top of active fuel. If the lo'Nest on scale reading is above the top of active fuel , then a reactor vessel level value should not be included. For plants that have implemented Westinghouse Owners Group Emergency Response Guidelines , enter the parameters used in the Core Cooling Red Path. Site specific indication of an inability to adequately remove heat from the RCS: [BWR] Use the Heat Capacit)' Temperature Limit. This addresses the inability to remove heat via the main condenser and the suppression pool due to high pool water temperature.
[PWR] Insert site specific parameters associated with inadequate RCS heat removal via the steam generators. These parameters should be identical to those used for the Inadequate Heat Removal threshold Fuel Clad Barrier Potential Loss 2.B and threshold RCS Barrier Potential Loss 2.A in the PWR EAL Fission Product Barrier Table. EGL Assignment Attributes:
3.1.3.B 267 SS8 ECL: Site Area EmergeRcy Initiating Condition:
Loss of all Vital DC power for 15 miRutes or loRger. --------------~ 1%1 99 GI (ReYisioA
: 6) November 2G 12 Of)erating Mode Af)f)lieability:
Power OperatioR , Startup , Hot 8taRdb;*, Hot 8hutdovm1, 2, 3, 4 Examf)le Emergeney Aetion Levels: Note: The EmergeRcy Director should declare the Site Area EmergeRcy promptly upoR determiRiRg that 15 miRutes has beeR e)weeded, or will likely be e)weeded.
l lAdicated voltage is less thaR (site specific bus voltage value)l 15 VOC OR ALL (site specific Vital DC busses) 1(2) D O l, D 02, D 03, aRd D 04 for 15 miRutes or loRger. Basis+ SAFETY 8Y8TEM: /*, system required for safe plaRt operatioR, cooliRg dowR the plaRt aRd/or placiRg it iR the cold shutdovm coRditioR, iRcludiRg the EGGS. Systems classified as safety related. This IC addresses a loss of Vital DC power which compromises the ability to moRitor aRd coRtrol SAFETY 8Y8TEM8. IA modes above Cold 8hutdowR , this coRditioR iRvolves a major failure of plaRt fuRctioRs Reeded for the protectioR of the public. FifteeR miRutes was s~lected as a threshold to e,wlude traRsieRt or momentary power losses. EscalatioR of the emergeRcy classificatioR level 1 Nould be via ICs AGlB.Ql , FGl or 808. DeYelof)er Notes: The "site specific bus Yoltage value" should be based OR the miRimum bus voltage Recessary for adequate operatioR of SAFETY 8Y8TEM equipmeRt.
This voltage value should iRcorporate a margiR of at least 15 miRutes of operatioR before the oRset of iRability to operate those loads. This voltage is usually Rear the miRimum voltage selected wheR battery siziRg is performed. The typical value for aR eRtire battery set is appFO>(imately 105 VDC. For a 60 cell striRg of batteries , the cell voltage is apprmdmately 1.75 Volts per cell. For a 58 striRg battery set , the miRimum voltage is approximately 1.81 Volts per cell. The " site specific Vital DC busses" are the DC busses that provide moRitoriRg aRd coRtrol capabilities for SAFETY 8Y8TEM8. EGL AssigRmeRt Attributes:
3.1.3.B 268 1'IBI 99 O I (ReYisioR e) NoYemeer 2012 SG1 ECL: General Emergency Initiating Condition:
Prolonged loss of al-l-ALL offsite and ALLalt onsite AC power to emergency essential buses. Operating Mode Applicability
: Power Operation , Startup , Hot Standby , Hot ShutdovmL..1.,__J Example Emergency Action Level s: Note: The Emergency Director should declare the General Emergencyevent promptly upon determining that (site specific hours) the applicable time 4 hours has been exceeded, or will likely be exceeded. a. Loss of ALL offsite and ALL onsite AC power to 1A3 and 1 A4 specific emergency buses). AND b. EITHER of the following:
Definitions:
_* _Restoration of at least one AC emergency essential bus in less than specific hours)4 hours is not likely. OR * (Site specific indication of an inability to adequately remo*,e heat from the serejRPV level cannot be restored and maintained above -25 inches. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems classified as safety related. Basis: Th i s IC addresses a prolonged loss of all power sources to AC emergency essential buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/press u re control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers.
In addition , fission product barrier monitoring capabi l ities may be degraded under these cond i tions. The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG 1. This will allow additional time for implementation of offsite protective actions. Escalation of the emergency classification from Site Area Emergency will occur if it is projected that power cannot be restored to at least one AC emergency essential bus by the end of the analyzed 4 hour station blackout coping period. Beyond this time , plant responses and event trajectory are subject to greater uncertainty, and there is an increased l i kelihood of challenges to multiple fission product barriers.
269 
~JEI 99 QI (ReYisioA e) ~Je*remeer 2Q 12 The estimate for restoring at least one essentialemergency bus should be based on a realistic appraisal of the situation.
Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for , and implement , protective actions for the public. 270 
 
NEI 99 0 I (RevisioA
: 6) No&#xa5;ember 2012 The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core. De*,reloper Notes: Although this JG and EAL may be vie*,*,ced as redundant to the Fission Product Barrier IGs , it is included to provide for a more timely escalation of the emergency classification level. The " site specific emergency buses" are the buses fed by offsite or emergency AG pov,*er sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typical I)' 1 emergency bus per train of SAFETY SYSTEMS. The "site specific hours" to restore AG power to an emergency bus should be based on the station blackout coping analysis performed in accordance with 10 GFR &sect; 50.63 and Regulatory Guide 1.155 , Stetien Bkwkeut. Site specific indication of an inability to adequately remove heat from the core: [BWR] Reactor vessel ;vater level cannot be restored and maintained above Minimum Steam Cooling RPV \1/ater Level (as described in the EOP bases). [PWR] Insert site specific values for an incore/core e>dt thermocouple temperature and/or reactor vessel water le*,rel that drive entry into a core cooling restoration procedure (or othervrise requires implementation of prompt restoration actions).
Alternately , a site may use incore/core e>(it thermocouple temperatures greater than l , 200&deg;F and/or a reactor vessel water level that corresponds to appro>(imately the middle of active fuel. Plants with reactor vessel level instrumentation that cannot measure down to approximately the middle of active fuel should use the lowest on scale reading that is not above the top of active fuel. If the lowest on scale reading is above the top of active fuel , then a reactor vessel level value should not be included.
For plants that have implemented Westinghouse Ovmers Group Emergency Response Guidelines , enter the parameters used in the Gore Cooling Red Path. EGL Assignment Attributes:
3.1.4.B 272 NEI 99 0 I (ReYisioR
: 6) r>Jovemeer 2012 SG8SG2 ECL: General Emergency Initiating Condition:
Loss of al-I-ALL AC and Vital DC power sources for 15 minutes or longer. Operating Mode Applicability:
Power Operation , Startup , Hot Standby , Hot Shutdown.l.,_l,_l Example Emergency Action Level s: Note: The Emergency Director should declare the General Emergencyevent promptly upon determining that the applicable time 15 minutes has been exceeded , or will likely be exceeded.
: a. b. Definitions:
Loss of ALL offsite and ALL onsite AC power to (site specific emergency
---b-u ... s-e-s)+-1A3 and 1A4 buses for 15_-minutes or longer. AND Indicated voltage is less than (site specific bus voltage value) 105 VDC on Abb (site speeifie Vital DC busses) BOTH Div 1 and Div 2 125 VDC buses for 15 minutes or longer. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related.A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems elassified as safety related. Basis: This IC addresses a concurrent and prolonged loss of both AC and Vital DC power. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling , containment heat removal/pressure control , spent fuel heat removal and the ultimate heat sink. A loss of Vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both AC and DC power will lead to multiple challenges to fission product barriers.
Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met. De,*eleper Netes: The "site specific emergency buses" are the buses fed by offsite or emergency AC pov;er sources that supply power to the electrical distribution system that powers SAFETY SYSTEMS. There is typically l emergency bus per train of SAFETY SYSTEMS. The " site specific bus voltage value" should be based on the minimum bus voltage necessar)' for adequate operation of SAFETY SYSTEM equipment.
This Yoltage value should incorporate a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This voltage is usually near the minimum voltage selected when battery sizing is performed.
273 NE[ 99 0 I (Re*,*isieA a) Nevemeer 20 I 2 The typical value for an entire battery set is apprmlimately l 05 VDC. For a 60 cell string of batteries , the cell voltag e is approximately 1.75 Volts per cell. For a 58 string battery set , the minimum voltage is apprmlimately 1.81 Volts per cell. The " site specific Vital DC busses" are the DC busses that pro11ide monitoring and control capabilities for SAFETY SYSTEMS. This IC and EAL 1.vere added to Revision 6 to address operating experience from the March , 2011 accident at Fukushima Daiichi. EGL Assignment Attributes:
3.1.4.B 2 7 4 
}ffil 99 0 I (Re\*isioA e) }fo\1 eme er 20 1 2 APPENDIX A -ACRONYMS AND ABBREVIATIONS AC ......................................................
..........................
...................................... Alternating Current AOP ...............................................................................
..................
Abnormal Operating Procedure A .......................................................................................................................................... r, PFJi.4 ........................................................
............................................ /\verage Po 1.ver Range Meter ATWS ........................
...........................................................
Anticipated Transient Without Scram .......................................................................................................................................... 8 &&#xa5;.' ..........................................................
......................................................... Babcock and \l/ilcox ............................................
.............................................................................................. B HT ....................................................................
................... Boron Injection Initiation Temperature BWR ...............................................................
.....................
......................... Boiling Water Reactor CDE ................................................................................
......................
Committed Dose Eq uivalent CFR .........................
........................
..............................
.................
...... Code of Fe deral Re g ul at ions CT~4T/CNMT .................................................................
.............................................. Containment
.......................................................................................................................................... C 8F ....................................................
................
............................
................ Critica l Safety Fu n ction .......................................................................................................................................... C 8F8T ...........................................................
.......................
...... Critical Safety Function Status Tree ..............................................................................
............................................................ D BA .....................................................................
............................................ Design Basis i\ccident DC ...................................................
................................
....................
....................... Direct Curre nt EAL .......................
...................................
..........................................
....... E mergency Action Level ECCS ..................
...............................................
..................
.........
E mer ge nc y Core Cooling System ECL ....................................
......................
.........................
.............
E mer ge nc y Class i fication Level E OF ...................
..........................................
......................
............... E mergenc y Operations Fac ilit y E OP .....................
.....................................
..................
...................
Eme r gency Operating Procedure EPA ......................................................................................
....... Env iron1nental Protection Agency E PG ...........................................................
........................
............
E mer ge nc y Procedure Guideline
..............................................................................
............................................................ E PIP .......................................................
..........................
.. gmergency Plan Implementing Procedure
.......................................................................................................................................... E PR ...............................
.......................................................
................... g yolutionary Po\ver Reactor .......................................................................................................................................... E PRJ ....................................................
.................................
.......... Electric Power Research Institute
.......................................................................................................................................... g RG ...................................................................................
............... Emergency Response Guide l ine .......................................................................................................................................... F EMA .................
..............
...........................
......................
Federal Emergency Management Agency f8AR .......................................................
............................................ fin al Safety Ana l ysis Report GE ...........
................................
...............
......................................
...................... General E mergency HC+L ...................................................................
....................... Heat Capacity Temperature Lim i t HPCI .............
....................
.............................................
................
High Pressure Coo l ant Injection
...............................................
........................................................................................... H 81 .......................................................................
..........
............................... Human System Interface IC .....................................................................................
...............
....................
lnitiating Condition Nel 99 01 (Re\'i s ioA e) T>lo&#xa5;efl'leer 2012 ...................................................
.......................................................
................
................ I D ...............
.....................................................................
...........................................
Inside Diameter IP E EE. ............................ Individual Plant E>rnmination of External Events (Generic Letter 88 20) ISFSI ...........................................................................
Independent Spent F u el Storage Installation Keff ....................................................................................
Effective Neutron M ultiplic ation Factor LCO ...............................
.............................
..........................
......... Limiting Conditio n of Operation
...................................................
................
.......................................................................
L OCA .........................................................................................
................. Loss of Coo l ant Accident ****************************************************************************************************************************************** ~4 CR ................................................
.................................
.........................
........... ~4ain Control Room ..................................
..............
.....................
.........................
.....................
.....................
.. ~4 SIV .............
..............
.........................................
.............
....................... ~4ain Steam Isolation Valve ~<ISL ...............
.........................
........................................
................
....................... ~<fain Steam Line mR , mRem , mrem , mREM ....................
................................
........ milli-Roentgen E quivalent Man MW .......................
.................
..................
......................
....................................................
Megawatt NEI ................................
.............................................................................
Nuclear E ner gy Inst itut e ...............
.............
...............................................................................
................
...............
l>+ PP ...........
................
...............................
..........
.............................................
.... l'+uclear Po\ver Plant .......................................................................................................................................... N RC ................................
.......................
.............
...............
..............
N ucl ear Regulatory Commission NSSS ...............................................................
..................................
Nuclear Steam Supply System NORAD ................................
..........................
....... North American Aerospace Defense Com mand fN O)UE ..........................................................
...........
.....................
fN ot ifi cation O ff Unusual Event NUMARC 1 *************************
*********************************
***** N ucl ear Management and Resources Co un ci l OBE ....................................................
............................
..................
..... Operating Basis Eart hquake OCA .......................
....................
...............
....................
..................
............. Owner Co ntr olled Area .................................
..............
.....................
....................
...................
............................... 0 f)GML ODAM .........................................................
Offsite Dose Calculation (Assessment) Man u al ORO .........................
........................
.................................
..............
Off site Re s ponse Organiz:ation PA ...........
.....................
...............
............
............................................................
....... Protected Area ...................
............
.....................
.....................................
................................................. p ACS .........................
....................
.........................................
Priority Actuation and Centro I System PAG ........................
..................................
........................
...........
.......... Protective Action Guide lin e *************************
**************************************************************************
*******************
******************** p JCS ................................................
................................... Proces s Information and Control System PRA/PSA ....................................
Probabilistic Risk Assess m e nt/ Probabilistic Safety Assessment PWR ........................................................................................................
Pressurized Water Reactor ******************
***************
***********************************
***************************
******************************************* p S ..........................................................
.................
..................
.............................. Protection System PSIG ........................................................................................
......... Pounds per Square Inch Gauge R .......................................
...................
.................
....................
.......................................... Roentgen ........................................................................
.................................................................. R CC ...........................................................
...........
....................
..........
.......... Reactor Control Console RCIC .....................
.....................................
.....................................
Reactor Core Isolation Coo lin g 1 NUMARC was a predecess or o r ga nization of the Nuc l ear Energy In st itut e (NE I). A-2 NE! 99 0 I (Re&#xa5;isioA
: 6) ~io&#xa5;ember 2012 RCS .........................................................................
.......................
............. Reactor Coolant System Rem, rem , REM .....................................................................
................. Roentgen Equivalent Man .........................
...................
.............
.................
................
.............................................
... R ETS ......................
....................
............
.............
...... Radiological Effluent Technical Specifications RPS ................................
................
...............................
.......................... Reactor Protection System RPV ............
....................
...............
.................................
.................
............ Reactor Pressure Vessel ..............................
.................
...........................................................................................
R VLIS ...................
.......................................
...........
.... Reactor Vessel Level Instrumentation System RWCU ......................................................................................
.................... Reactor Water Cleanup ..........................................................................................................................
................ s AR ....................................
......................
............
...............................
........... Safet)' ,<\nalysis Report ***********************************************************
****************************************
******************
*********************
s AS .......................
........................
...........
.....................
................
........... Safety ,<\utomation System .......................
...........................
......................
............................
......................................
s BO .................................................
..............
.............................
..........................
..... Station Blackout SCBA ...............
...................................
................................... Self-Contained Breathing Apparatus
..............................
...............
................
...................
.....................................
..................... s G .........................
.........................
...................................
................
......................... Steam Generator
............................................................................
...................................................
...........
s I ..........................................
...................
...........................
........................................ Safety Injection
............................
...............................................
....................................
.........................
.. s JCS ....................................................................
................. Safety Information and Control System ******************************
**********************************************
********************
******************************************
s PDS ...................
..........................................................
................. Safety Parameter Display System SRO .................
................................................
...........................................
Senior Reactor Operator TEDE .................................
.........................
...................................
Tota l Effective Dose Equivalent T G AF ............................................................
..............................
........................ Top of Active Fue l TSC ..................
...................................
................
..................................... Technical Support Center ........................
.............................................
..............................
.............
.......................... U FSAR ............................
.....................
...................................
Updated Final Safety Analysis Report \VOG ...................
..........................
.....................................................
\Vestinghouse 0 1.vners Group NEI 99 QI (ReYisioA a) November 2Ql2 APPENDIX B -DEFINITIONS t>i:El 99 0 I (Re.,.isioA
: 6) }fo.,.ember 2012 The following definitions are taken from T itl e 10 , Code of Federal Regulations, and related regu lat ory guidance documents.
Alert: Events are in progress or have occ urr ed which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels. General Emergency: Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actua l lo ss of physical contro l of the facility. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Notification of Unusual Event (NOUE)\ Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systemsSAFETY SYSTEMS occurs. Site Area Emergency:
Events are in progress or have occurred which involve actua l or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that resu lt s in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to , equipment needed for the protection of the public. _Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.
The following are key terms necessary for overall understanding the NEI 99 01DAEC emergency classification scheme. Emergency Actio n Level (EAL): A pre-determined , site-specific, observable threshold for a n Initiating Condition that, when met or exceeded , places the plant in a g i ven emergency classification level. Emergency Classification Level (ECL): One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditio n s according to (1) potential or actual effects or consequences , and (2) resulting onsite and offsite response actions. The emergency classification leve l s , in ascending order of severity , are: Notification of Unusual Event (NOUE) Alert Site Area Emergency (SAE) General Emergency (GE) + This term is sometimes shorteAed to UAusual E.,.eAt (VE) or other similar site speeifie termiAology.
B-1 
~lEI 99 01 (Re\*isioR
: 6) November 2012 Fission Product Barrier Threshold:
A pre-determined, site-specific, observable threshold indicating the loss or potential loss of a fission product barrier. Initiating Condition (IC): An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.
Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters ( e.g., ALL CAPS). These words are defined terms that have specific meanings as used in this document.
The definitions of these terms are provided below. CONFINEMENT BOUNDARY: (Insert a site speeific definition for this term.) Developer Note -The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. This corresponds to the pressure boundary for the Dry Shielded Canister (DSC) shell (including the inner bottom cover plate) base metal and associated confinement boundary welds. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.
For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.
CONTAINMilff CLOSURE: (Insert a site speeific definition for this term.) DeYeloper Note Site speeific proeedurally defined aetions taken to seeure eontainment and its assoeiated struetures, systems, and eomponents as a funetional barrier to fission product release under e>(isting plant eonditions.
For DAECs, this is considered to be Secondary Containment as required by Teehnieal Specifieations.The proeedurally defined eonditions or aetions taken to secure containment (primary or secondary for BWR) and its assoeiated struetures , systems , and eomponents as a functional barrier to fission product release under shutdovm eonditions.
DESIGN BASIS EARTHQUAKE (DBE): A DBE is vibratory ground motion for which certain (generally, safety-related) structures, systems, and components must be designed to remain functional.
EXPLOSION:
A rapid , violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding , arcing , etc.) should not automatically be considered an explosion.
Such events may require a post-event inspection to determine if the attributes of an explosion are present. F/\ULTED:
The term applied to a steam generator that has a steam leak on the seeondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steam generator to beeome eompletely depressurized.
Developer Note This term is applicable to PV/Rs only. B-2 
-----------------------------------------------------
Ne! 99 0 I (RevisioR
: 6) ~/0 1 ,emaer 2012 FIRE: Combustion characterized by heat and li ght. Sources of smoke such as slipping drive belts or overheated electrica l eq uipm ent do not constitute FIRES. Observation of flame is preferred but is NOT required if l arge quantities of smoke and heat a r e observed. HOSTAGE: A person(s) held as l everage aga in st the stat i on to ensure that demands will be met by the station. HOSTILE ACTION: An act toward a NP-llnuclear power plant or its personnel that includes the use of violent force to destroy equipment , take HOSTAGES , and/or intimidate the lic ensee to ac hi eve an end._ This includes attack by air , land, or water using guns , explos i ves , PROJECTILEs , veh i cles , or other devices used to de li ver destructive force. _Other acts that sat isf y the overa ll int ent may be inc luded. HOSTILE ACTION shou ld not be construed to include acts of civil disobedience or fe l on iou s acts that are not part of a concerted attack on the NPJ!nuclear power plant._ Non-terrorism-b ased EALs should be used to address such activities (i.e.,_-this may include violent acts between individuals in the owner controlled area) .
* HOSTILE FORCE: One or more individuals who are engaged in a determined assau l t , overtly or by stea lth and deception , equ ipp ed with su it ab l e weapo n s capable of killing , maiming , or causing destruction.
IMMINENT:
The trajectory of events or conditions is s u ch that an EAL will be met within a re lativel y short period of time regardless of mitigation or corrective actions. INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI): A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. NOR~4AL LEVELS: As applied to radiologieal IG/EALs , the highest reading in the past twenty four hours exeluding the eurrent peak value. OPERA TING BASIS EARTHQUAKE (OBE): An OBE is vibratory ground motion for which those features of a nuclear power plant necessary for continued operat i on without undue risk to the health and safety of the public will remain functional.
OWNER CONTROLLED AREA: (Insert a site speeifie definition for this term.) Developer Note This term is typically taken to mean the site property owned by , or otherw i se und er the co ntrol of , the licensee.
Tn some cases , it may be appropriate for a lieensee to define a smaller area with a perimeter eloser to the plant Proteeted Area perimeter (e.g., a site with a large OGA where some portions of the boundary may be a signifieant distanee from the Proteeted Area). In these eases , developers should eonsider using the boundary defined by the Restricted or Seeured Ovmer Controlled Area (ROGfJSOGA). The area and boundar)'
seleeted for seheme use must be eonsistent 1 ,vith the deseription of the same area and boundary eontained in the Seeurity Plan. PROJECTILE:
An object directed toward a NP-llnuclear power plant that could cause concern for its continued operability , rel i abi lit y, or personnel safety. B-3 NE! 99 QI (RevisieA e) ~Jevemaer 2Q 12 PROTECTED AREA:_ (Insert a site specific definition for this term.) De~relaper Nate This term is typically taken to mean ti he area under continuous access monitoring and control , and armed protection as described in the site Security Plan. REFUELING PATHWAY:_ (Insert a site specific definition for this term.) Develaper Nate This description should include all the cavities, tubes, canals and pools through which irradiated fuel may be moved , but not including the reactor vessel. Includes all the cavities, tubes, canals and pools through which irradiated fuel may be moved, but not including the reactor vessel. RtJPTURE(D):
The condition of a steam generator in 1 , 1 ,rhich primar)' to secondary leakage is of sufficient magnitude to require a safety injection. Develaper Nate This term is applicable to PWRs only. SAFETY SYSTEM: A system required for safe plant operation , cooling down the plant and/or placing it in the cold shutdown condition , including the ECCS. These are typically systems are classified as safety-related.
Develaper Nate This term may be modified to include the attributes of " safety related" in accordance with 10 CFR 50.2 or other site specific terminology, if desired. SECURITY CONDITION:
Any Security Event as listed in the approved security contingency plan that constitutes a threat/compromise to site security , threat/risk to site personnel , or a potential degradation to the level of safety of the plant. _A SECURITY CONDITION does not involve a HOSTILE ACTION. SITE BOUNDARY:
That line beyond which the land is neither owned, nor leased. nor otherwise controlled by the Company. UFSAR Figure 1.2-1 identifies the DAEC SITE BOUNDARY.-: UNISOLABLE:
An open or breached system line that cannot be isolated , remotely or locally. 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. 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.
Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. B-4 B-5 NEI 99 g I (Re'lisieA e) NeYember 2012 NEI 99 0 I (Re,*isioA
: 6) ~tOY6fl'l06F 20 J 2 APPENDIX C PERMANENTLY DEFUElED STATION ICs/EAls Recognition Category PD provides a stand alone set of ICs/EALs for a Permanently Defueled nuclear power plant to consider for use in developing a site specific emergency classification scheme. For development , it was assumed that the plant had operated under a 10 CFR &sect; 50 license and that the operating company has permanently ceased plant operations.
Further , the company intends to store the spent fuel within the plant for some period of time. When in a permanently defueled condition , the plant licensee typically receives approval from the NRG for eJ(emption from specific emergency planning requirements.
These eJ(emptions reflect the lowered radiological source term and risks associated 1.vith spent fuel pool storage relative to reactor at power operation.
Source terms and accident analyses associated
\.Vith plausible accidents are documented in the station's Final Safety Anal)1 sis Report (FSAR), as updated. As a result, each licensee will need to develop a site specific emergency classification scheme using the :NRG approved exemptions , revised source terms , and revised accident analyses as documented in the station's FSAR. Recognition Category PD uses the same ECLs as operating reactors; however , the source term and accident analyses typically limit the ECLs to an Unusual Event and Alert. The Unusual Event ICs provide for an increased awareness of abnormal conditions 1.vhile the Alert ICs are specific to actual or potential impacts to spent fuel. The source terms and release motive forces associated with a permanently defueled plant would not be sufficient to require declaration of a Site Area Emergency or General Emergency.
A permanently defueled station is essentially a spent fuel storage facility 1.vith the spent fuel is stored in a pool of water that serves as both a cooling medium (i.e., removal of decay heat) and shield from direct radiation. These primary functions of the spent fuel storage pee I are th e focus of the Recognition Category PD ICs and EALs. Radiological effluent IC and EALs were included to provide a basis for classifying eyents that cannot be readily classified based on an obserYable e\.1 ents or plant conditions alone. Appropriate ICs and EALs from Recognition Categories A, C , F , H , and S 1 ,\.1 ere modified and included in Recognition Category PD to address a spectrum of the events that may affect a spent fuel pool. The Recognition CategOF)'
PD ICs and EALs reflect the relevant guidance in Section 3 of this document (e.g., the importance of avoiding both over classification and under classification).
Nonetheless , each licensee will need to develop their emergency classification scheme using the NRG approved eJ(emptions , and the source terms and accident analyses specific to the licensee.
Security related events will also need to be considered.
C-1 l>lel 99 0 I (Re1t*isioR
: 6) l>Jo,*emeer 2012 Table PD 1: Reeognition Category "PD" Initiating Condition MatFix UNUSUAL EVENT PD ,A .. Ul Release of gaseous or liquid radioactivit)*
greater than 2 tirnes the (site specific effluent release controlling docurnent) limits for 60 rninutes or longer. Op. },lodes: llfetApplicehlc PD AU2 UNPLA1'J1'ffiD rise in plant radiation levels. Op. A fede s: Ne tA ppli c ehle PD SUl UNPLA1'J1'ffiD spent fuel pool temperature rise. Op. },lodes: l*let Applicehlc PD HUl Confirrned SECURITY C01'JDITI0N or threat. Op. },lodes: ,VetAppliceblc PD HU2 Hazardous event affecting SAFETY SYSTEM equipment necessary for spent fuel cooling. Op. Modes: ,"hlet Appliceblc PD HUJ Other conditions e>dst which in the judgment of the Emergency Director *.varrant declaration of a (NO)UE. Op. },fedes: ,"hlet Applicehlc ALERT PD AAl Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrern TEDE or 50 rnrern thyroid CDE. Op. },fedes: ,"hletAppliceble PD AA2 UNPLA1'J1'mD rise in plant rad i at i on le;*els that irnpedes plant access required to maintain spent fuel integrity.
Op. },fedes: , Vet Applirnble PDHAl HOSTIL E ACTION v,ithin the OW1'J:ER CONTROLLED AREA or airborne attack threat within 30 minutes. Op. },fedes: A'etApplicehle PD HAJ Other conditions e>dst which in the judgrnent of the Ernergency Director warrant declaration of an Alert. Op. },{edes: ,VetAppliceble
,-------------------, : Table intenaea for use by 1 1 BAL aevelopers.
: Inclu s ion in licensee C-2 I ..I , , ..I , uocurnents 1s not requ1reu. 1 L------------------J 
---~------------ECL: Notification of Unusual Event ~/EI 99 Ql (Revision
: 6) ~lovomeor 2Ql 2 PD AU1 Initiating Cenditien:
Release of gaseous or liquid radioactivit)
' greater than 2 times the (site specific effluent release controlling document) limits for 60 minutes or longer. Netes: Operating Mede Applieability:
Not Applicable Example Emergeney Aetien Levels: (l or 2)
* The Emergency Director should declare the Unusual Event promptl)' 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 unlmown , assume that the release duration has e>weeded 60 minutes.
* If the effluent flow past an effluent monitor is knovm to have stopped due to actions to isolate the release path , then the effluent monitor reading is no longer valid for classification purposes. (l) (2) Reading on ,A_._..""JY effluent radiation monitor greater than 2 times the alarm setpoint established by a current radioactiYity discharge permit for 60 minutes or longer. Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the (site specific effluent release controlling document) 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 e>weeds regulatory commitments for an extended period of time (e.g., an uncontrolled release).
It includes any gaseous or liquid radiological release , monitored or un monitored , including those for which a radioactivity discharge permit is normally prepared.
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 e>(tended , 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 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. C-1 NEI 99 o I (Re*,*isieR e) Jl,levemeer 2012 ti:! te the mes that a release pa ed *ter readiAgs assu . . l'Ae"'A te haYe stepp 8 ffiHOR! !HORI ffi t lflORl!eFIS
' j" dfeF ClassifieatieR
_8*s: ~~:e efflHeRt flew past aR * "":,er rea8iRg is "" leRge, va , ,, AmeAt is estabhs!:ie . at!:i tl=ieA the efflueAt me eA~ ire . . late the release p ' 8HO !0 80!l0RS le 199 8' g 4 times *c: f A purpeses.
I a release exoee IA elassiNea-,e-
-8 Fe, &deg;'""'P**, Aet be prerated er averag; . Releases sheHlf AHies Sees Rel !fleet the llnL. a<liatieR ffi0Riter release liffiits ie, 39"' e* aetivity releases that eause h~:*;::lflit.
+his eAL B' L #1 +his on . . lal,lishe8 By a ro ,0 ' tiRH0HS release pa-" c A :b addresses ra 10 d. aot 11*1ty d1so ar t!:i.,*ays n 2
* s the l1m1t es ffem nen readiAgs te e>,oeed t1~e~ .. :ti:! planAed bato!:i releases "
* iealiy Be as,eeiate n I 8 B " w1ll l}jl ,.. te gas). . that are Seteete ) (e.g., ra8waste, ..as eR!rnlle8 gaseeus er liq~,e ~'::~::,.ys (e.g., spills ef . g, L addresses""" Rlfl0R1te,e ) B* L #2 +his~ I r>'eys paFtieularly 0R * . . *er>**atersystelfls , ete .. , r
* eAta su-. ' I I e IA rn sample aAalyses er eAv1reAm draiAs heat e>wl=iaAgerea rng ra<lieaetive liquids iRte stem, ' . . ... 18 Be via IC PD AA!. esoalatien ef the emer geAoy olassifioatieA le,,el weu N t s* n. h Ra<lielegieal Effl**Rt D""eleper e;,
* eeRtrnlliRg 8ee~ffieRt 19 t
* C:eRerie Leiter 89 91 , +he "site Sfl**ifie effiueRt ,el~aseplaRts that have 11F1pie1HeRte8 le1HeRt "'gulatieRs related . Ii ati eRs (Rll+S) **, <<>* MJ +hese 8eeulfleRts
'"'P. l ' *l'JJrnp,iate, the +eehRieal Si**: ~aleulatieR MaRual (O~C;;.9 ~FR Part 59 , AppeR81K Lt;,:shtles ier this IC. the Offs1te es E 10 CFR Part 2 aA t blisl=iiAg the memter 11 1 eeRtrnls e.g., 18 B *see Jer es* M te ef HOR he8elegy sheu -e -. 1Uff8 er 00C~. RB+8 er ODCM !flat *10,s 8eseriBe8 IR the
* e the offiueRt ffieRI . I . RiteFS sh0Hl8 iRelu e . te8 with ether pete*ti* Listed lfl0 . . t lied lfl0RiteFS ass*~** e e B' L values ier i8er i**iu81Rg IRS a {;M"H If IRelu * , fl e . Develepers
"'"Y al:::~::eriBe8 i* tRe llll+8 er~~.;;,~
Sesokelease lilflit:,v:*::::*
tR>>*ays tRat are
* the 1H0st "PP 1 , Be Belew """ efflueR! pa--.. R 18 Be 8eterlfliAe8 usmg *-I lated BAL value "'a) : R r t Alse , selfle these Fll0Rll0FS s *:i It is reeegaiaee tRat a ea **.1 ROOS 10 Be iRelue.ee IR t * .::.d ,elated the llll'.f8 e, ,?~Ci~ ;Rat ease, the lfl0~te, a~":;e:i!ieatieRs e, ethe~ ~**.:s:.::/eR elearly i8eRtify lfl0R1te, eaR ' .. med By +ee ***
* 1 e BAL aR '"'' meRiteFS "'"Y *et B: ge,i: is i1Hp0FtaRt that the assee,~~FS . ts* t!:iere43re , bT . f tliese ffi0R
* 1 reqH1re1HeR--, --h seer avaihH ,ty e . . I sos with sepa,a-e a*y li1Hitati0*s
** t *
* address gaseeus aR8 hqu,e re** 8eme sites m8) . , fiAd it advaAtageeus te u ,..,,,...,...,,.,, ., ~7 {on *'*he 'Feehn,eo.~
{; ' l*o/ P,og,*""'
' I** so,Ha*(*)
wh"* , __ ,. '""''""''"""'
-~*1 ."'; ',,;;:,,.1
..
* 10 /he P,*~-*""'',; *** ;. 1h, ,;,. .... ,.. ** , ' ; ', Ojfeile Do,e c.,"'"'&deg;''":'" flh effi,,at memlo,s '" .. *a O By l>JPO ,.1a1, l,re ........ 8 0 h o ***** , ,,.., O Thi, iAol,Oa, ";~**~FR >o.47(1,)(8)
,.; (9). I ofl O CFR SO .S4 (q) .,; H ~:**ilo,s.
the reE!HiremeRtS e J iR miRe the FeE!HIF~lfle~
S h aeeitieR efether efffoeR H De**elepers sheule ceep . Rt , .. heR eeRs1eermg t e . ro * ..,, . .,..., ,e ' . 'S eifieoi,on, ** ' . h e; ss . , $ 'fiec1tief'ls il'I .'he . , . Bl Ejfh:e:?t Teehme~, ~e;t;:, edt!rnl Detc1i!s &j RET. le ti C,ntrelsfer R86He,egie_ . dtfie Releec1ltel'I 8.rr. ee
* EfHipme geRey respeRse e te emer C-2 NEI 99 0 I (Re*,isioA e) Jl,lo*,emeer 2012 Radiati_on mo.niter readings should reflect va l ues that correspond to a radiolo ical release exceedmg 2_times a releas_e ~ontrol limit. The controlling document typical I)~ describes
~eetho:olo~ies fo~ dete~mmmg effluent radiat i on monitor setpoints; these methodologies should .~se to etermme En~ values. In cases where a methodology is not adequate! , defined :***1';1'"" skeelB Beterm1Ae valees eeAsisteAt with effleeAt eeAtrel ,egelatieAs
(/g lQ cf~ art and IO CFR Part 50 Appendix T) and related guidance.
* ., . ~o: EA_L #1 Values in this EAL should be 2 times the setpoint established b , the ::::**elivity 8 15 ei,arge P*Fffi it le wam ef * ,el ease tkat is Aot iA eompl iaAee with tke ;peeifieB ts: Inde)ong the *~alue m this manner ensures consistency betv~*een the EAL and th t
* established by a speeific discharge permit. e se13omt Developers should researeh radiati~n monitor design documents or other information s~uree~ to ensure th~t I) the EAL value bemg considered is within the usable response and displ_a) rang~ of ~he 1~strument , and 2) there are no automatic features that may render the momtor readmg mvahd (e.g., an auto purge feature triggered at a particular indication level). ffl It is recognized that the condition described by this IC may result in a radio l ogical e uen~ :*~ue ~eyond the operating or display range of the installed effluent monitor In those case~~ 1' ;.a u~s sho~ld be determined with a margin sufficient to ensure that an ac.curate mom-or rea-mg 1s a"a1lable Fe I c AL *
* f .* ..r ex.amp e , an en momtor readmg might be set at 90% to 95% o~~he h1g7est accurate m~mtor re~dmg. This provision notwithstanding , if the est1matedrcalculated monitor read mg is greater than approx i mately 110% of the h" h t *t d. h ' 1g es accurate ~oni:~r rea mg , t en developers may choose not to include the monitor as an indication and 1 ent1 'an alternate EAL threshold. , _Indications from a real _time do~~ projection system are not included in the eneric EALs t~aA) heeR;ees Bo AO! ka*,e th!S eapah,hty.
eor those that Bo, tke eapal,ility may A:t ~e ***itkiA * --e scope o-the p l t T h
* I 8 *fi
* 1-~ 1 f d~n :ec mcapec1 cations. A licensee may request to inelude an EAL using rea ,me ose prOJectJon s ystem results; approval ',&#xa5;ill be considered on a case by case ;asis. C f; 4 /ndications from a perimeter monitoring system are not included in the generic EA Ls
* any 1censees d~ no~ have this capability.
for those that do , these monitors may not
* T:::o~le~
:nd ;amt~med to the ~~me I eve~ as plant equipment , or within the scope of the plant mca 6 ~eCI cations. In add1t10n , readmgs may be influenced by environmental or other actor~. 1 1 1:.~1censee m_ay request to include an EAL using a perimeter monitoring system; a ppro1-a n ii I be considered on a case by case basis. EGL Assignment Attributes:
3.1.l.B C-3 I-8 "cVV Od: JO l \IV 0d :)l l'l!A eq pf ROA\ f9A9f UO!ll'lO!J!SSBfO
,(ouei5J9UJ9 94lj0 UO!ll'JfBOSff "Sf 13!J9ll3W 9lSl3A\ 9A!l0l'lO!Pl'lJ JO weweAOUJ pua S90JROS 0!4de~O!PBJ JO esn se 4ons se9!A!lOl'l peuue1d WOJJ l[RSeJ ll'l4l seseeJOU!
teAet UO!ltJ!PBJ sepn1oxe c# 1\/!I *doJp 1eAet J9ll3M G!IN:N:V1d:Nfl ua Ol enp S! i5U!pl'leJ pelBAete e4l eJe4"A sestJo LI! ,(1uo e1qeo!Jdde S! t# 1\/'3 ll34l 9lON *peJep!suoo eq p1no4s suO!lRfOAe peuutJjdJO Sl99JJ9 e4+/- "SUO!ll'JOOJ eS04l U! SJOl!UOW ,(q pel09l9P eq UB9 lB4l Sl'JeJe lUeoefpl'lJO Sf9A9f UO!ll'l!Pl'lJ e41 U! 9Sl'J9J0U!
ue esneo OSfl'l Al'JUJ f9A9f J9ll3A\ e41 LI! doJp lUl39!J!Ui5!S y *Ee1qef!BABJ!)
suO!ll'JAJesqo tJJeweo oep!A JO f9UU0SJed lUl3fd UJOJj s&#xb5;odeJ epRfOU! , (l'JUJ SUO!ll'lO!PU!
f0A9f JO S09JROS J94l0 "UO!ll'JlU0WR.llSU!
f9A9f e1qef !l'lAl'l UJOJJ suO!lBO!PU!
,(q peu!UJJelep
, (f!JBW!Jd eq lf!A*, estJeJoep f9A9f J9ll'J,'A y "lUl3Jd 04lj0 Al9jl3S JO J9A0J 04l U! UO!lBPBJi50p J13!lU9lOd 13 S! UO!l!PUOO J94l!3 " SJ13!J9ll3UJ 9A!l0130!Pl'lJ JO lUl3Jd e41 U!4l!N. SJ9A9f UO!ll3!Pl3J fOJlUOO Ol ,(l!f!ql'l e41 U! sso1 JOU!lli 13JO 9A!ll'l0!PU!
eJe s1eAet UO!lB!P13J peseeJOU!
e4+/- *slueAe G'3N:N:V1d:Nfl Je410 JO 1enJ E1ueds) petB!Ptl"!
eAoqtJ 1eAe1 JeltJNr LI! estJeJoep tJ ,(q pesntJo SfeAef UO!ll'l!Pl'lJ lUBfd peltJAete sesseJppl'l
:)I S!4+/- "81!IA311Vt"q1f0N JQAO J4ffllli &sect;cJO es!J G3NNV1d:Nfl ue sell'lO!PU!
lfRSeJ , (e,uns JO i5u!peeJ JOl!UOUJ UO!lll!Pl'lJ tJeJv "SJOl!UOW UO!ll'l!PtlJ i3U!MOjf Oj e4uo _;\~TV' , (q pell'JO!PU!
Sl'J SJ9A9f UO!ll'l!Pl'lJ l'J9Jl'J U! 9S!J G3NNV1dN:fl
:i5U!A\OIIOJ e4uo :A.NV ,(q pell'lO!PU!
stJ 100d 1eRJ 1ueds e4l LI! doJp 1eAet Je1e;A G3:N:N:V1d:Nfl "t) Ee) Et) lU9Aff jl3RSRUf1 JO UO!lBO!J!lON_
:'}.)';I znv ad z l oz JaqwaAON:
(~ IW!S!Aa'd)
LO 66 13:!"1:
Develof)eF Notes: NEI 99 Ql (Re\'isioR e) "!>Jove me er 2Q 12 For EAL #1 Site specific indications may include instrumentation values such as water level and area radiation monitor readings , and personnel reports. If available , video cameras may allow for remote observation.
Depending on available in strumentation , the declaration may also be based on indications of water makeup rate and/or decreases in the lev el of a water storage taflb For EAL #2 The specified value of 25 mR/hr may be set to another value for a specific application with appropriate justification.
EGL Assignment Attributes:
3.1.1.B B-2 
~/El 99 QI (ReYisioR
: 6) November 2012 ECL: Notification of Unusua:I;, E:*,e=n~t
--------------~PRO SU 1 Initiating Cenditien:
UNPLA1'H-ffi9 spent fuel pool temperature rise OpeFating Mede A I' . .
* af)f) iealllh*"*
Not A 1* hi ~J
* 1tpp-1cat1-e Example EmeFgeney Aetien Levels: (1) UNPLANNE9 spent fuel I pee temperature rise to greater than (s1*te s *.c: o peClt1C Fr-This IC addresses a condition that. potential degradation in the level o~ s:f:te~ursor to a more serious event and represents
: a. occur , and result in a loss of pool level a:d ~fthe plant. T~u.ncorrected , soiling in the pool will . mcreased rad1at1on levels. Escalation of the emer e . g nc~* classification level *would ee via IC pg A A 1 >> .. eloper Notes: ' ** PD AA2. Th
* e site specific temperature should e EGL A
* n.ss1gnment Attrieutes:
3 .1.1.A B-3 ECL: Notification of Unusual Event Initiating Cenditien:
Confirmed SECURITY CONDITION or threat. Operating Mede ,\pplieability:
Not Applicable Example Emergeney Aetien Le~1 els: (I or 2 or 3) NEI 99 QI (ReYisieA
: 6) l>le&#xa5;ember 2Ql 2 PD HU1 (1) A SECURITY CONDITiffi,1 that does not involi,re a HOSTILE ACTION as reported by the (site specific security shift superi,rision).
(2) (3) }ofotification of a credible security threat directed at the site. A validated notification from the NRG providing information of an aircraft threat. Thi_s I~ addre~ses events that pose a threat to plant personnel or the equipment necessary to mamtam cool_mg of spent ~el , and thus represent a potential degradation in the level of plant safet_y. Security events v,rh1ch do not meet one of these EALs are adequately addressed by the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72. Security events assessed as HOSTILE ACTiffi,lS are classifiable under IC PD HAl. Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security related event. Classification of these events will initiate appropriate threat related notifications to plant personnel and OROs. Security plans and terminology are based on the guidance provided by NEI 03 12 , Temptatcfor the Security Ptan , Treining e,9d QHelificetion Plen , Se.fcgMerds Contingency Plen {end Independent Spent-Fuel St-orege Instell8tio1~
Security Progrmn}.
EAL #1 references (site specific security shift supervision) because these are the individuals trained to confirm that a security event is occurring or has occurred.
Training on security event confirmation and classification is controlled due to the nature of Safeguards and 1 O CFR &sect; 2.39 information. EAL #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with (site specific procedure).
EAL#~ addresses the threat from the impact of an aircraft on the plant. The NRG Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft.
The status and size of the plane may also be provided by NORAD through the NRG. Validation of the threat is performed in accordance with (site specific procedure). Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate gecurity sensitive information.
This includes information that may be
* B-4 NEI 99 QI (ReYisieA
: 6) ~Jeyemeer 2Q 12 advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
Security sensitive information should be contained in non public documents such as the Security Plan. Escalation of the emergency classification le*,el would be via JC PD HAI. DevelepeF Netes: The (site specific security shift supervision) is the title of the on shift individual responsible for supervision of the on shift security force. The (site specific procedure) is the procedure(s) used by Control Room and/or Security personnel to determine if a security threat is credible , and to validate receipt of aircraft threat information.
Emergency plans and implementing procedures are public documents; therefore , EALs should not incorporate Security sensitive information.
This includes information that may be advantageous to a potential adversary , such as the particulars concerning a specific threat or threat location.
Security sensitive information should be contained in non public documents such as the Security Plan. With due consideration given to the above developer note , EALs may contain alpha or numbered references to selected events described in the Security Plan and associated implementing procedures.
Such references should not contain a recognizable description of the event. For e>rnmple , an EAL may be 't't'Orded as " Security event #2 , #5 or #9 is reported by the (site specific security shift superYision)." EGL Assignment Attributes:
: 3. I. I .A B-5 Nel 99 o I (ReYisioH G) No,*ember 2012 PD HU2 ECL: Notification of Unusual Event Initiating Cenditien:
Hazardous event affecting SAFHTY SYSTEM equipment necessary for spent fuel cooling. Operating Mede Applieability:
Not Appl i cable Example Emergeney A,etien Le,;els: (]) a. b. C. The occurrence of i+ ...... ~Y of the following hazardous events:
* Seismic event (earthquake)
* Seismic event (earthquake)
* Internal or external flooding event
* Internal or external flooding event
* High winds or tornado strike
* High winds or tornado strike
* FIRE
* FIRE
* EXPLOSION  
* EXPLOSION (site specific hazards)River level above 757 feet
* (site specific hazards)
* River Water Supply (RWS) pit low level alarm
* Other events vrith similar hazard characteristics as determined by the Shift Manager The event has damaged at least one train of a SAFETY SYSTEM needed for spent fuel cooling. AND The damaged SAFETY SYSTEM train(s) cannot , or potentially cannot , perform its design function based on EITHER:
* Other events *with similar hazard characteristics as determined by the Shift Manager or Emergency Director ---AND b. EITHER of the following:
* Indications of degraded performance
101 NEl 99 01 (Re\'isioR 6) }loYeFRber 2012 ------1. Event damage has caused indications of degraded performance in-at least one train of a SAFETY SYSTEM needed for the current operating mode. 2. 2EITHER of the following:.,-
* VISIBLE; DAM.AGE This IC addresses a hazardous event that causes damage to at least one tra i n of a SAFHTY SYSTEM needed for spent fuel cooling. The damage must be of sufficient magnitude that the system(s) train cannot , or potentially cannot , perform its design function. This cond i tion reduces the margin to a loss or potential loss of the fuel clad barrier , and therefore represents a potentia l degradation of the level of safety of the plant. For HAL l.c , the first bullet addresses damage to a SAFHTY SYSTHM train that is in service/operation since indications for it will be readily available.
* Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode,-ef, _* _The event has caused resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM component or structure needed for the current operating mode.::;: -. 102 Definitions: "t-ffiI 99 01 (ReYisioR
For EAL l.c , the second bullet addresses damage to a SAFETY SYSTHM train that is not in service/operation or readily apparent through indications a l one. Operators will make this B-6 NE[ 99 0 I (Re*,*isieA
: 6) "t-l0\'8fl'!88F 2012 FIRE: Combustion characterized by heat and light. Sources of smoke such a s 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.
: 6) Jl,fo*,remeer 2012 determination based on the total it)' of available event and damage report information.
EXPLOSION:
This is intended to be a brief assessment not requiring lengthy anal)'Sis or quantification of the damage. Escalation of the emergency classification level could , depending upon the event , be based on any ofthe Alert ICs; PD AAl , PD AA2 , PD HAl orPD HA3. Devel01Jer Netes: For (site specific hazards), developers should consider including other significant , site specific hazards to the bulleted list contained in EAL 1.a (e.g., a seiche). Nuclear power plant SAFETY 8Y8TEM8 are comprised oftvro or more separate and redundant trains of equipment in accordance with site specific design criteria.
A rapid, violent and catastrophic failure of a piece of equipment due to combustion.
chemical reaction.
or overpressurization.
A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an e x plosion are present. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related
.A system required for safe plant operation.
cooling down the plant and/or placing it in the cold shutdovm condition.
including the EGGS. Systems classified as safety related. 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.
Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of d e graded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential e x ists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria CA6. l .b. l of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. An event affecting a single-train SAFETY SYSTEM (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under CA6 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train SAFETY SYSTEM, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement.
103 Nel 99 QI (Re*,isioR e) No&#xa5;ember 2() 12 Indications of degraded performance 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. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. 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 f*ftAh EAL l .b. l 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 sigAificant 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 eyent 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 AS1RS1. De:velepeF Netes: For (site specific hazards), developers should consider including other sigAificant , site specific hazards to the bulleted list contained in EAL l .a (e.g., a seiche). Nuclear po 1 n<er plant SAFETY SYSTEMS are comprised of two or more separate and redundant trains of equipment in accordance 1.vith site specific design criteria.
EGL Assignment Attributes:
3.1.2.B 104 NEI 99 QI (Re,,*isioR a) ~lo 1 rem0er 2012 CS1 ECL: Site Area Emergency Initiating Condition:
Loss of (reactor vessel/RCS
[PWR] or RPV [BWR]) inventory affecting core decay heat removal capability.
Operating Mode Applicability:
Cold Shutdown , Refuelin~
Emergency Action Levels: Example Emergeney Aetion LeYels: (1 or 2 or3) Note: The Emergency Director should declare the Site Area Emergencyevent promptly upon determining that the applicable time 30 minutes has been exceeded , or will likely be exceeded.
~1 a. b. ~2 a. b. CONTAINMENT CLOSURE not established. AND (Reactor vessel/RCS
[PWR] or RPV [BWR]) level LESS THANless than specific level)+64 inches" CONTAINMENT CLOSURE established.
AND (Reactor vessel/RCS
[PWR] or RPV [BWR]) level LESS THA1'J:less than specific level).+ 15.:.: inches ~3 a. (Reactor vessel/RCS
[PWR] or RPV [BWR]) level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by AN&#xa5;-EITHER of the following:
I Definitions:
* (Site.33 ecific radiation monitor) Drywell Monitor (9184A/B) reading greater than (site :33 ecific value)5.0 R/hr
* Erratic source range monitor indication
[PWR]
* UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool UNPLANNED increase in (site specific sump and/or tank)_levels of sufficient magnitude to indicate core uncovery * (Other site specific indications) 105 tl-EI 99 QI (ReYisioR
: 6) tlovemaer 2012 CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.
For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.
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. 106 Basis: NE! 99 01 (ReYisioA
: 6) NoYemeer 2012 This IC addresses a significant and prolonged loss of (reactor vessel/RCS
[PWR] or RPV [BWR]) inventory control and makeup capability leading to IMMINENT fuel dama g e. 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.
107 NET 99 01 (ReYisioA e) November 2012 Following an extended loss of core decay heat removal and inventory makeup , decay heat will cause reactor coolant boilin g and a further reduction in reactor vessel level. If RG-8,l reactor vessel level cannot be restored , fuel damage is probable.
Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINM E NT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RG-8,l reactor vessel levels of EALs CS 1.1.b and CS 1.2.b reflect the fact that with CONTAINMENT CLOSURE established , there is a lower probability of a fission product releas e to the environment.
-.. In the Cold Shutdown and Refueling Modes, LT/LI-4559 , 4560 , and 4561 (RX VESSEL NARROW RANGE LEV E L) instruments read up to 22" high due to hot calibrations.
LI-4541 (WR GEMAC , FLOODUP) should be used in these Modes for comparison to EAL thresholds s ince it is calibrated cold and reads accurately.
If normal means of RPV level indication are not available due to plant evolutions, redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted.
In EAL CS 1.3.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.
The inabilit y to monitor (reactor vessel/RCS
[PWR] or RPV [BW~]) 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
[PWR] or RPV [BWR]). These EALs address concerns raised by Generic Letter 88-17, Lo s s of Deca y Heat Removal; SECY 91-283 , Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Pow e r Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guid e lin es for Industry Actions to Assess Shutdown Management.
---Escalation of the emergency classification level would be via IC CGI or AG+RGl. De, 1 el0f)er Notes: Accident analyses suggest that fuel damage may occur vrithin one hour of uncovery depending upon the amount of time since shutdown; refer to Generic Letter 88 17, SECY 91 283, NUREG 1449 and NUMARC 91 06. The type and range of RCS level instrumentation may vary during an outage as the plant moves through various operating modes and refueling evolutions, particularly for a PWR. As 108 tl-el 99 01 (ReYisioR
: 6) NoYember 2012 appropriate to the plant design , alternate means of determining RCS level are installed to assure that the ability to monitor level 1 Nithin the range required by operating procedures will not be interrupted.
The instrumentation range necessary to support implementation of operating procedures in the Cold Shutdo\lm and Refueling modes may be different (e.g., narro\1,'er) than that required during modes higher than Cold Shutdovm.
For EAL #1.b the " site specific level" is 6" below the bottom ID of the RCS loop. This is the le 1 1el at 6" below the bottom ID of the reactor vessel penetration and not the low point of the loop. If the availability of on scale level indication is such that this leYel value can be determined during some shutd01, 1 ,'n modes or conditions , but not others , then specify the mode dependent and/or configuration states during 1tvhich the level indication is applicable.
If the design and operation of water leYel instrumentation is such that this level value cannot be determined at any time during Cold Shutdovm or Refueling modes , then do not include EAL #1 (classification will be accomplished in accordance 1 ,vith EAL #3). For EAL #2.b The "site specific level" should be apprmdmately the top of active fuel. If the availability of on scale le 1 ,rel indication is such that this level value can be determined during some shutdown modes or conditions , but not others, then specify the mode dependent and/or configuration states during 1 ,vhich the level indication is applicable.
If the design and operation of water level instrumentation is such that this level value cannot be determined at any time during Cold Shutdovm or Refueling modes , then do not include EAL #2 (classification will be accomplished in accordance with EAL #3). For EAL #3 .b first bullet As water level in the reactor vessel lowers , the dose rate above the core will increase.
Enter a " site specific radiation monitor" that could be used to detect core uncovery and the associated
" site specific value" indicative of core uncovery.
It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases , EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available.
For e>rnmple, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold.
To further promote accurate classification , developers should consider if some combination of monitors could be specified in the EAL to bui Id in an appropriate level of corroboration between monitor readings into the classification assessment.
For EAL #3.b second bullet Post TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations.
For EAL #3.b third bullet Enter any 'site specific sump and/or tank" levels that could be e>(pected to change if there were a loss of RCS/reactor vessel inventory of sufficient magnitude to indicate core uncovery.
Specific level values may be included if desired. For EAL #3.b fourth bullet Developer s should determine if other reliable indicators exist to identify fuel uncovery (e.g., remote viewing using cameras). The goal is to identify any unique or 109 NEI 99 0 I (Re1t*isioR 6) No;remeer 2012 site specific indications , not already used else*where , that 1tvill promote timely and accurate emergency classification.
For EAL #1.b " site specific level" is the LoY,r Low Low EGGS actuation setpoint / Level 1. The BWR Lov,* Lov,r Low EGGS actuation setpoint / Level 1 was chosen because it is a standard operationally significant setpoint at which some (typical!)
' low pre s sure EGGS) injection systems would automatically start and attempt to restore R.0 V level. This is a RPV water level value that is observable below the Lov,r Low/Level 2 value specified in IC CAI , but significantly above the Top of Active Fuel (TOAF) threshold specified in EAL #2. For EAL #2.b The " site specific level" should be for the top of active fuel. For EAL #3 .b first bullet As *;,rater level in the reactor vessel lowers , the dose rate above the core will increase.
Enter a " site specific radiation monitor" that could be used to detect core uncovery and the associated
" site specific value" indicative of core uncovery.
It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases , EAL values should be determined 1.Yith a margin sufficient to ensure that an accurate monitor reading is available.
For e>rnmple , an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold. To further promote accurate classification , developers should consider if some combination of monitors could be specified in the EAL to build in an appropriate level of corroboration between monitor readings into the classification assessment.
For BWRs that do not have in s talled radiation monitors capable of indicating core uncovery , alt e rnate site specific level indications of cor e uncovery should b e used if available.
For EAL #3 .b second bullet Because BWR source range monitor (SRM) nuclear instrumentation detectors are typically located below core mid plane , this may not be a viable indicator of core uncovery for BWRs. for EAL #3.b third bullet Enter any " site specific sump and/or tank" levels that could be e>,pected to change if there were a loss ofRPV inventory of sufficient magnitude to indicate core uncovery. Specific level values may be included if desired. For EAL #3.b fourth bullet Developers should determine if other reliable indicators e1dst to identify fuel unco 1 ,1ery (e.g., remote viewing using cameras).
The goal is to identify any unique or site specific indications , not already used elsewhere , that will promote timely and accurate emergency classification.
EGL Assignment Attributes:
EGL Assignment Attributes:
3 .1.1.A and 3 .1.1 C B-7 NEI 99 0 I (RevisioA
3 .1.3 .B 110 J!,ffil 99 g l (ReYisioR
: 6) November 2012 PD HU3 ECL: Notification of Unusual Event Initiating Condition:
: 6) NoYemeer 2Q 12 CG1 ECL: General Emergency Initiating Condition:
Other conditions e>(ist which in the judgment of the Emergency Director warrant declaration of a (NO)UE. Operating Mode Applieability:
Loss of (reactor vessel/RCS
Not Applicable Example Emergeney Aetion Levels: (1) Other conditions exist v,hich in the judgment of the Emergency Director indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated.
[PWR] or RPV [BWR]) inventory affecting fuel clad integrity with containment challenged.
No releases of radioactive material requiring offsite response or monitoring are e>(pected unless further degradation of safety systems occurs. This IC addresses unanticipated conditions not addressed e>(plicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a NOUE. B-8 ECL: Alert Nel 99 GI (RevisioR
Operating Mode Applicability:
: 6) J>Jo,*ember 2G 12 PD AA1 Initiating Condition:
Cold Shutdown , Refueling4. 5 Emergency Action Levels: Examf)le EmergeneyEmergeney Aetion Levels: (1 or 2) Note: The Emergency Director should declare the General Emergencyevent promptly upon determining that the applicable time 30 minutes has been exceeded , or will likely be exceeded.
Release of gaseous or liquid radioactivit)*
C 1.1 a. (Reactor vessel/RCS
resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Operating Mode Applieability:
[PWR] or RPV [BWR]) level LESS THA1'tless than fs+te-specific level)+ 15 !!inches for 30 minutes or longer. C 1.2 AND bl!. ANY indication from the Secondary Containment Challenge Table (see below)C-1. a. (Reactor ve ss el/RCS [PWR] or RPV [BWR]) level cannot be monitored for 30 minutes or longer. -----AND b. Core uncovery is indicated by EIHERAN&#xa5; of the following:
Not Applicable Example Emergeney Aetion Le*;els: (1 or 2 or 3 or 4) Notes:
* Drywel I Monitor (9184A/B) (Site specific radiation monitor) reading GREATER THANgreater than (site specific value)5.0 R/hr.
* The ?merg~ncy Director should declare the Alert promptly upon determining that the applicable time has been e>(ceeded , or *Nill likely be exceeded.
* Erratic source range monitor indication
* If an ongoin~ release is detected and the release start time is unknown , assume that the release duration has exceeded 15 minutes. * ~f the effluent flovr past an effluent monitor is known to haYe stopped due to actions to isolate the release path , then the effluent monitor reading is no longer *ralid for classification purposes.
[PWR]
* The pre calculated effluent monitor Yalues presented in EAL #l should be used for emergency classific~ion assessments until the results from a dose assessment us i ng actual meteorology are aYa1Jable.
* UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool illiPLA1'J1'JED increase in (site specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery ,. * (Other site specific indications)
(1) (3) (4) Read~ng on ANY of the following radiation monitors greater than the reading shown for 15 mmutes or longer: (site specific monitor list and threshold values) (2) Dose assessment
AND 111 Nel 99 Q l (Re*,isioR e) No*,ember 2Q 12 c. ANY indication from the Secondary Containment Challenge Table (see below~ l}. Table C-1 Secondary Containment Challenge Table
~sing actual meteorology indicates doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site specific dose receptor point). " I . f 1*
* CONTAINMENT CLOSURE not established*
* n:na y~1s o a1qu1d effluent sample indicates a concentration or release rate that would resul_t m doses greater than 10 mrem TEDE or 50 mrem thyroid CDE at or beyond (site specific dose receptor po int) for one hour of e>(posure.
* Drywell Hydrogen or Torus Hydrogen GREATER THA}lgreater than 6% AND Drywell Oxygen or Torus Oxygen GREATER THANgreater than 5% (Explosive mixture) exists inside containment
Field surve~ results indicate EITHER of the follovring at or beyond (site specific dose receptor point):
* UNPLANNED increase in containment pressure
* Closed windov, dose rates greater than 10 mR/hr e>(pected to continue for 60 minutes or longer.
* Secondary containment radiation monitors above max safe operating limits (MSOL) of EOP 3, Table 6radiation monitor reading above (site specific value) [BWR]
* Analyse_s of fie.Id survey samples indicate thyroid CDE greater than 50 mrem for one hour of mhalat1on.
* If CONTAINMENT CLOSURE is re-established prior to exceeding the 3 0-_minute time limit , then declaration of a General Emergency is not required.
Basis: This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual B-9 NEI 99 QI (Re&#xa5;isioR e) November 2Q 12 ef fsite deses greater thaR er equal te 1% efthe BPA Preteetive Aet~eR Gui~esd(PAGs).
112 Definitions: "t>IEI 99 Q 1 (ReYisioR e) "t>lo1~emeer 2Q 12 CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.
1! eludes beth meRitered aRd uR meRitered releases. Releases ef this magmtu _e r~preseR , aR iR ae tual er peteRtial substaRtial degradatieR efthe level ef s~fi:ty efthe pl~Rt_as md1::!eedR~?o1~ed dielegieal release that sigRifieaRtly e , weeds regulatery l1m1ts (e.g., a s1gmfieaRt ra re lease). R adiele ieal efflueRt EALs are alse iReluded te previde a basis for ela_ssifyiRg eveRts _a~d aRElitia~s that eaRRat be "'&<lily ar app_ropriate~y
For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.CONTAINMENT CLOSURE: The procedurally defiAed 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.
*~*s:ifia<I
UNPLANNED:
~" ;;e.::*~*, ~::::::: 1~~1~8":,esses e al eRe. The iRelusieR ef beth plaRt eeRd1tl8R aR r~ _ie eg1ea e u 1' t he speetrum ef pessible aeeideRt eveRts aRd eeRd1tl8RS.
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.
h TeDE d
The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged.
* tat 1% efthe EPA PAG ef 1 , 000 mrem 1 Nhile the 50 mrem thy_reid T e ese is se . A AG :t; TeDE Rd thyre1d COE was established iR eeRsideratieR efthe 1 :5 ratie ef the EPn P,, ~r a Gf)B-: e ClassifieatieR based eR efflueRt meRiter readiRgs assumes that a _release path te the d *
This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity.
* bl" h d lf the efflueRt flev,c past aR efflueRt meRiter 1s lrnevm te have ~teppe d:: 1;:::~::~st:~!:i~tse
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 RGS1 reactor vessel level cannot be restored , fuel damage is probable.
:h~ release path , theR the efflueRt meRiter readiRg is Re leRger valid fer elassifieatieR purpeses.
With CONTAINMENT CLOSURE not established , there is a high potential for a direct and unmonitored release of radioactivity to the environment.
Developer Notes: While this JC may Ret be met abseRt ehalleRges te the eeeliRg ef speRt fuel , it pre~i~s elassifie~ieR diversity aRd may be used te elassify eveRts that *weuld Ret reaeh the same based eR plaRt eeRditieRs aleRe. The EPA PAGs are e>tpressed iR terms ef the sum efthe effe~tive des~ equivaleRt (EDE) eemmitted effeetive dese equivaleRt (CEDE), er as the thyre1d eemm1tted
If CONTAINMENT CLOSURE is established prior to exceeding the 30-minute time limit , then declaration of a General Emergency is not required.
~ese ::i~a~eRt (CDH). Far the perpase afthese ICIBALs , the Elase ~~aRtity ta~~~:~~*~'k
The existence of an explosive mixture means , at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (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.
.... ". equivaleRt (TEDE), as defiRed iR 10 CFR &sect; 20 , 1s used m lieu ef ... sum e The EPA PAG guidaRee prevides fer the use adul_t thyreid ~ese eew~ersieR faeters; he"'e"er seme states haYe deeided te base preteetive aetwRs eR eh_Ild thyre1d CD~. Nuelear , ., v 1' t IC LE, A Ls Reed te be eeRsisteRt with the preteetive aet1eR methedeleg1es emple) ed pewer p-aR s , n
It therefore represents a challenge to Containment integrity.
* h IC d EAL sheuld be b , the States withiR their EPZs. The thyreid COE des~ used i_R _t e a~, s. a~justed as Reeessary te aligR *Nith State preteetive aetwR dee1swR makmg entena. The " site speeifie meRiter list aRd thresheld values" sheuld be determiRed with eeRsideratieR ef the follewiRg:
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.
* SeleetieR ef the apprepriate iRstalled gaseeus aRd liquid effl~et~ meRitc;~DE er 50 mrem
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.
* The efflueRt meRiter r e adiRgs sheuld eerrespeRd te a dese mr~m . thyreid COE at the " site speeifie dese reeepter peiRt" (eeRs1steRt with the ealeulat1eR methedelegy empleyed) fer eRe heur ef expesure. .
In EAL CG 1.2.~, 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).
* MeRiter readiRgs will be ealeulated usiRg a set ef assumed meteereleg1eal data er ~tmespherie dispersieR faeters; the data er faeters seleeted for use sheuld be the same as these empleyed te ealeulate the meRiter readiRgs fer TC PD AUl
It also allows sufficient time 113 NEI 99 QI (ReYisioA
* B-10 Nm 99 01 (Re\*isioA e) Noyember 2012
: 6) J>loYeA'leer 2Q 12 for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.
* The calculation of monitor readings will also require use of an assumed release isotopic mi>E:; the selected mi>E: should be the same as that employed to calculate monitor readings for TC PD AUi.
For EAL CGl.2.b, the calculated radiation level on the Drywell Monitors (9184A/B) is without the reactor head in place. Calculated in radiation levels with the reactor head in place are below the normal variation in background readings of these monitors.
* Depending upon the methodology used to calculate the EAL values , there may be overlap of some values between different ICs. Developers will need to address this overlap by adjusting these values in a manner that ensures a logical escalation in the EGL. The " site specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish bet>.veen on site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan , and the procedural methodology used to determine offsite doses and Protecfr, , e Action Recommendations. The variation in selected dose receptor points means there may be some differences in the distance from the release point to the calculated dose point from site to site. Developers should research radiation monitor design documents or other information sources to ensure that 1) the EAL value being considered is within the usable response and display range of the instrument , and 2) there are no automatic features that may render the monitor reading invalid (e.g., an auto purge feature triggered at a particular indication level). It is recognized that the condition described by this IC may result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. In those cases , EAL values should be determined 1,vith a margin sufficient to ensure that an accurate monitor reading is available.
114 
For e>rnmple , an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision nohvithstanding, if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold.
~ffil 99 01 (Re,*isioR 6) ~lo't'ember 2012 The in ability to monitor (reactor vessel/RCS
Although the IC references TEDE , field survey results are generally available only as a " whole body" dose rate. For this reason , the field sur*,ey EAL specifies a " closed window" survey reading. Indications from a real time dose projection system are not included in the generic EALs. Many licensees do not have this capability.
[PWR] or RPV [BWR]) l eve l may be caused by instrumentation and/or power failures , or water level dropping below the range of ava il able instrumentation.
For those that do , the capability may not be within the scope of the plant Technical Specifications.
If water level cannot be monitored , operators may determine that an inventory loss i s occurring by observing c h anges in s ump and/or tank l evels. S ump and/or tank l eve l changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS
A licensee may request to include an EAL using real time dose projection system results; approval will be considered on a case by case basis. Indications from a perimeter monitoring system are not included in the generic EA.Ls. Many licensees do not have this capability.
[PWR] or RPV [BWR]). F or the Containment Challenge Table, Secondary Containment ma x safe operating (MSOL) limits from EOP 3 are defined as the highest parameter value at which neither: (]) equipment necessary for the safe shutdown of the plant will fail nor (2) personnel access necessary for the safe s hutdown of the plant will be preclud e d. +I h ese 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 Actions to Assess Shutdown Management. Develeper Netes: Accident analyses suggest that fuel damage may occur v,ithin one hour of uncovery depending upon the amount of time since shutdo,*rn; refer to Generic Letter 88 17 , SECY 91 283, NUREG 1449 and NUMARC 91 06. +he type and range of RCS level instrumentation may vary during an outage as the plant moves through various operating modes and refueling evolutions , particularly for a P\VR. As appropriate to the plant design, alternate means of determining RCS level are installed to assure that the ability to monitor level 'Nithin the range required by operating procedures 1 n<ill not be interrupted.  
For those that do, these monitors may not be controlled and maintained to the same level as plant equipment , or 1 ,vithin the scope of the plant Technical Specifications.
+he instrumentation range necessary to support implementation of operating procedures in the Cold Shutdown and Refueling mod e s may be different (e.g., narrower) than that required during modes higher than Cold Shutdown.
In addition , readings may be influenced by environmental or other factors. A licensee may request to include an EAL using a perimeter monitoring S)'Stem; approval 1 Nill be considered on a case by case basis. EGL Assignment Attributes:
For EAL #1.a +he " site specific level" should be approximately the top of active fuel. If the availability of on scale level indication is such that this level value can be determined during some shutdown modes or conditions , but not others , then specify the mode dependent and/or configuration states during which the level indication is applicable.
3.1.2.C B-11 
If the design and operation of 'Nater level instrumentation is such that this level value cannot be determined at any time during Cold Shutdovm or Refueling modes, then do not include EAL #1 (classification will be accomplished in accordance with EAL #2). For EAL #2.b first bullet As water level in the reactor vessel lowers , the dose rate above the core will increase.
*uO!ltlO!.J!lSRf 0ltl!JdOJddtJ 4l!'t\ UO!ltlO!(ddu 9!J!90ds u JOj 0R(UA J04l0UU Ol l0S 0q AUW J4fflW 00 I JO 0R(UA P0!J!00dS 04:+/- *&#xb5;ef'y' Utl Utl4l lUtl9!J!U~!S 0JOW ,(((tl!lU0lOd lU0A0 Utl S0!(dW! UO!ltlJRP
Enter a " site specific radiation monitor" that could be used to detect core uncovery and the associated
,(tlp Ot u SU 'i3U!i3UJ0AU lR04l!M 0J04 pesn S! 0RfUA 04l 's,(up Ot 04l J0AO pei3UJ0AU eq ueo 0R(tlA J4fflW s l 04l lU4l S0P!AOJd 'SJbl:OUl:Od:fl'lhO't[
" site specific value" indicative of core uncovery.
bl:19fd UfJ_l}O~'l}'VJ,f{)
It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases , EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available.
bl:fJ.1}190lf!,l19f:)  
For eJrnmple, an E AL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than 11 5 
'LtlO O'inlflN:JO r*ffrn uO!l908 4zlno4l(V
~ffil 99 01 (Re,*isioA 6) NoYeA'laer 2012 apprmdmately 110% of the highest accttrate monitor reading , then developers may choose not to inclttde the monitor as an indication and identify an alternate EAL threshold.
*s0W!l ,(ouudnooo peloedJ(e JOJ wewlsnfpu 4WA s,(up Or U! weJ 5 JO en(tJA 6 l ;)GD 04l WOJJ p0A!Jep S! J4fflWS I JO 0n1eA 04:+/- .lUU[d 04l JO ,(l9JUS JO f 0A0( 04l JO UO!lUPU.1:i30p (U!lUUlSqns (tl!lU0lOd JO (tlRlOtl 04l U! Sl(RS0J lU4l SS090U p0J!tldW!
To farther promote accttrate classification , developers shottld consider if some combination of monitors cottld be specified in the EAL to bttild in an appropriate level of corroboration between monitor readings into the classification assessment.
S!4l S! ll *sseooe lUtl(d AJtJSS090U U0lU0J4l AilUUO!J!Uzl!S Ol lU0!0!J.JRS S! AU!0P JO 00U0J0jJ0lU!
For BWRs that do not have installed radiation monitors capable of indicating core ttncovery, alternate site specific level indications of core ttncovery shottld be ttsed if available.
04l lU4l pep!AOJd 'i3U!J0J.10lU!
For EAL #2.b second bttllet Post TMI accident stttdies indicated that the installed PWR nttclear instrnmentation 1.vill operate erratically 1 Nhen the core is ttncovered and that this shottld be ttsed as a tool for making sttch determinations. Becattse BWR Sottrce Range Monitor (SR.M:) nttclear instrnmentation detectors are typically located belov,r core mid plane , this may not be a viable indicator of core ttncovery for BWRs. For EAL #2.b third bttllet Enter any " site specific sttmp and/or tank" levels that cottld be e>(pected to change if there were a loss of inventory of sttfficient magnitttde to indicate core ttncovery.
JO zlu!J0PU!4 sepRfOU! ,epedw!, '0J04 pesn sv *,(l!Jzl0lU!
Specific level valttes may be included if desired. For EAL #2.b fottrth bttllet Developers shottld determine if other reliable indicators e>(ist to identify foe! ttncovery (e.g., remote viewing ttsing cameras). The goal is to identif)' any ttniqtte or site specific indications, not already ttsed else 1.vhere , that will promote timely and accttrate emergency classification.
f0RJ lU0ds U!tllU!tlW Ol pepeeu SW0lS , (s U!tllU!tlW Ol J0pJO U! 'zlU!JOl!UOW (UOO( S0J!Rb0J lU4l JO A((URUUU:I p0lUJ0dO eq lSRW lU4l lU0Wd!nbe i3U!U!UlU09 seeJe Ol SS099tl AJtJSS000U epedw! lU4l S(0A0( UO!lU!PtlJ peseeJOU!
For the Containment Challenge Table: Site shtttdown contingency plans typically pro 1 ,ride for re establishing CONTAINMENT CLOSURE following a loss of RCS heat removal or inventory control fonctions.
S0SS0Jppe
For " E>(plosive mi>(tttre" , de1+*elopers may enter the minimttm containment atmospheric hydrogen concentration necessary to sttpport a hydrogen bttrn (i.e., the lower deflagration limit). A concttrrent containment 0>1.ygen concentration may be inclttded if the plant has this indication available in the Control Room. For BWRs , the use of secondary containment radiation monitors shottld provide indication of increased release that may be indicative of a challenge to secondary containment.
;)f 5!4+/- *,(l!lzl0lU!
The "site specific valtte" shottld be based on the E OP ma>1.imttm safe valttes becattse these valttes are easily recognizable and have a defined basis. EGL Assignment Attribtttes:
(0Rj lU0dS U!tllU!tlW Ol pepeeu SW0lS,(S JO UO!lUJedo JO (U!J0lUW 01.!lOUO!PUJ JO IOJlUOO U!UlU!UW Ol pepeeu seeJtJ zlu!A\.O((Oj 04uo X~v' Ol sseooe s0pedw! lU4l S'lili\'.3:9 9VJ,~ON: J0Ao J4fflW 001 ,(q 0S!J u 0lUO!PU! Sl(RS0J ,(01t1ns 10 szlu!pu01 JOl!UOJ,'{
3.1.4 .B 116 tlEI 99 QI (ReYisioR
UO!lU!PU1f ueJy G'.3:N:N: V1dN:fl :Al!Ji30lU!
: 6) }lo~*emeer 2Q 1 2 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS Table E 1: Reeognition Category "E" Initiating Condition Matrix UNUSUAL EVENT E HUl Damage to a loaded cask CONFINEMENT BOUNDARY.
(0RJ lU0dS U!UlU!UW Ol pepeeu SW0lS,(s JO UO!lUJedo JO (tl!J0lUW 0A!l0tlO!PtlJ JO (OJlUOO U!tllU!tlW Ol ,(ouednooo SRORU!lUOO i3U!l!Rb0J SU0JU i3U!MO()Oj 04lJO :A.NV U! 14mw s [ UU4l J0lU0.l:i3 0lUJ 0sop G'.3:N:N:V1dN:fl Ee) Et) *,(l!.l:i30lU!
Op. },ledes: All 117 I Table intenEleEI for use b)' 1 EAL EleYelopers.
(0Rj lU0dS U!tllU!tlW Ol p0l!Rb0J SS099U lUUJd sepedw! lU4l S(0A0( UO!lU!PUJ lUUJd U! 0S!J G'.3:N:N:V1dN:fl
: lnelusion in lieensee I ,.i * * ,.i 1 uoeurnents 1s not requ1reu.
:HO!J!f)HO.)
1 L------------------*
~H!JU!J!HI Z'f/'f/ Cd cl Ol JaqwaAON:
I8F8I MALFUNCTION ECL: Notification of Unusual Event Initiating Condition:
(~ lcl0!5!>'*0"M) t O 66 13:N:
Damage to a loaded cask CONFINEMENT BOUNDARY.
ECL: Alert Ne! 99 0 I (RevisioA e) ~lovemeer 2012 PD HA1 Initiating Condition:
Operating Mode Applicability:
HOSTILE i\CTION v,'ithin the OWNER CONTROLLED AREA or airborne attack threat 1.vithin 30 minutes. Operating Mode Applieability:
All Example Emergency Action Levels: E-HU1 E HUI .1 Dama ge to a loaded cask CONFINEMENT BOUNDARY as indicated by an on contact&#xa3;!
Not Applicable Example Emergeney Aetion Levels: (1 or 2) (1) (2) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the (site specific security shift supervision).  
radiation reading greater than the values shown belewon Table E-1 (2 times the site specific cask specific technical specification allowable radiation level) on the surface of #le-spe nt fuel cask. Table E-1 Cask Dose Rates 61BT DSC 3 feet from HSM Surface 800 mrem/hr Outside HSM Door-Centerline of DSC 200 mrem/hr End Shield Wall Exterior 40 mrem/hr Definition:
/*, validated notification from NRG of an aircraft attack threat 1 Nithin 30 minutes of the This IC addresses the occurrence of a HOSTILE ACTION within the O\l/NER CONTROLLED AREA or notification of an aircraft attack threat. This event v,ill require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA , or the need to prepare the plant and staff for a potential aircraft impact. Timely and accurate communications bet\Neen Security Shift Supervision and the Control Room is essential for proper classification of a security related event. Security plans and terminology are based on the guidance provided by NET 03 12 , Templ8tc fer the Security Plan , Trainil'lg end Qualificetion Plan , Sejegbu1rds Co1'1tingenc;*
CONFINEMENT BOUNDARY:
Plen {end Independent Spent Fuel Storege Instelletion Security Progreni}. As time and conditions allow , these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation , dispersal or sheltering).
The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a sto rage cask containing spent fuel. It applies to irradiated fuel that i s licen se d for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors , and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of " damage" is determined by radiological survey. The technical specification multiple of " 2 times", which is also used in Recognition Category A-R IC RA Ul , is used here to distinguish between non-emergency and emergency conditions.
The Alert declaration will also heighten the awareness of Offsite Response Organizations , allowing them to be better prepared should it be necessary to consider further actions. This IC does not apply to incidents that are accidental events , acts of civil disobedience , or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. E>rnmples include the crash of a small airoraft , shots from hunters , physical disputes between employees , etc. Reporting of these types of events is adequately addressed by other EA.Ls , or the requirements of 10 CFR &sect; 73.71 or 10 CFR &sect; 50.72. EAL #1 is applicable for any HOSTILE ACTION occurring , or that has occurred , in the O\V1'mR CONTROLLED AREA. This includes any action directed against an TSfST that is located within the OW1'mR CONTROLLED AREA. EAL #2 addresses the threat from the impact of an airoraft on the plant , and the anticipated arrival time is within 30 minutes. The intent of this EAL is to ensure that threat related B-13 NEI 99 0 I (RevisioA
_ The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the " on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSis are covered under !Cs HUI and HAL 118 61[ *0oua!1dwo3 JO 0ll30!J!'J.103 04l u! p0laoo1 UO!ll30!J!Oeds IUO!U400l S ,)fSUO 04i U! pUAOj eq uuo )(SUO !0RJ lU0ds u JOj f0A0! UO!fU!PUJ 01qaN,Ol(l3 04+/- &deg;("Ol0 '3)lfl0HDfV3
: 6) November 2012 notifications.
'31H.>f 'NOIS0'1dX 3 ')fSl30 J0AO pedd!l JO peddoJp u '*~*e) SlU0Ae epuw uaw JO IUJAll3U p0!J!lU0P!
are made in a timely manner so that plant personnel and OROs are in a heightened state of read mess. This EAL is met when the threat related information has been validated in accordance with (site specific procedure). !he NRG He~dquarters Operations Officer (HOO) 'Nill communicate to the licensee if the threat =*es an aircraft.
JO e~uaJ 04l WOJJ l(AS0J p1noo ll34l e~awap sesseJppa
The status and size of the plane may be provided by NOR/\.D through the In some cases , it may not be readily apparent if an aircraft impact 1.vithin the O\l/NER CONTR?LLED A~A ~vas intentional (i._e., a HOSTILE ACTION). It is e),pected , although not c_ertam , that not1fica~1on by an appropnate Federal agency to the site would clarify this point. In this case , the appropnate federal agency is intended to be NORAD , FBI , FAA or }iRC. The em~rgenc):
'IV3 S!lf+/- "kttVGtffiOff
?eclar~ion , _including one based on other ICs/EALs , should not be unduly delayed while awaitmg not1ficat10n by a Federal agency. Em:rgency plans and implementing procedures are public documents; therefore , EALs should no: , mcorporate Security ~ensitive information.
+/-N:3~\i:'3:NHN:03 04l l00JJU ,\l(U!lU0lOd P(AOO lU4l SUO!l!PUOO lU0P!OOl3 pua SlU0lr0 uu0wou04d ft3JAll3U 04l ,\J!lU0P! ':&#xb5;ode"tt UO!ll3A(l3A3 Al0Jl3S 3~N: pell3(0J 04l pua 0oua!1dwo3 JO 0ll30!J!'J.103
This includes information that may be ad" antageo~s to a pot:nt1al a~v.ers~ry , such as the particulars concerning a specific threat or threat location. Secunty sens1t1ve mformation should be contained in non public documents such as the Security Plan. Denio per Notes: The (si_t~ specific securi~y shift supervision) is the title of the on shift individual responsible for superv1s1on of the on shift security force. Em:rgenC)'
)(Sl3o 04l U! p0ou0J0j0J
plans and implementing procedures are public documents; therefore , EALs sh~, uld not mcorporate S:curity sensitive information. This includes information that may be ad rantageo~s to a pot:ntial a~v.ers~ry , such_as the particulars concerning a specific threat or threat location.
'ttVS u JO '[9&#xa3;&sect; I D~flN: .10d] ("ttVS) :&#xb5;ode~ S!SA(UUV Al0Jl3S JS.>fSI 04uo Sl(AS0J 04+/- N0I.L3NflA'IVJi\t ISASI 9 FISSION PRODUCT BARRIER ICS/EALS Table 9 f l: RecognitioR Category "f" IRitiatiRg CoRditioR Matrix ALERT A Ry Loss or aRy PoteRtial Loss of either the fuel Clad or RCS barrier. Loss or PoteRtial Loss of aRy two barriers.
Security sens1t1ve mformat1on should be contained in non public documents such as the Securit)' Plan. urth d *d * * ,r1ue cons1 erat1on given to the above developer note , EALs may contain alpha or ~umbered ~eferences to selected events described in the Security Plan and associated
Loss of aRy hNo barriers aRd Loss or PoteRtial Loss of the third barrier. See Table 9 F 2 fer :s,&#xa5;R EALs See Table 9 F 3 fer PWR EALs Jl,J:gJ 99 () 1 (Revision e) Jl,lo*remeer 2() 12 Devele~er Nate: The adjaceRt logic tlo\v diagram is for use by deYelopers aRd is Rot required for site specific im):)lemeRtatioR
-1mplementmg procedures.
; however , a site specific scheme must iAclude some t)'):)e of user aid to faci I itate timely aRd accurate classificatioA of fissioR product barrier losses aRd/or poteRtial losses. Such aids are typically com):)rised of logic tlov,r diagrams , " scoriRg" criteria or checkbox type matrices.
Such references should not contain a recognizable description of the e','.ent.
The user aid logic must be coRsisteRt with that of the adjaceRt diagram. 120 m I I ~I LOSS POTENTI A L LOSS FU E L CLAD LOSS POTENT I AL LOSS FUEL CLAD L OSS POTENT I A L LOSS FU EL CLAD LOSS LOSS LOSS 1/2 POT E NTIAL LOSS POTENTIAL L OSS RCS LOSS POTENTIAL LOSS Y ES [Gl -Lo ss of A N Y Two Barriers A.till. Lo ss or Potentia l Los s of Third Barrier POTENTIAL LOSS CON T A INMENT &#xa3;SJ. -Loss or Po t ential Loss of ANY Two Barriers ~--------------------
Fo_r exampl~, an EAL may be worded as " Security event #2 , #5 or #9 is reported by the *(site specific secunty shift supervision)." See the related Developer Note in Appendi>, B , Definitions , for guidance on the development of ,a scheme definition for the OWNER CONTROLLED AR~A. EGL Assignment Attributes:
--t,.i .EA.l-ANY Loss or A NY Potential Loss of filill.ER Fuel C lad Q& R CS 121 l'>J"EI 99 0 I (Re'>*isioR 6) 1'1e&#xa5;effieer 2012 N N .....;
3.1 .2.D B-14 NE! 99 Ql (ReYisioA 1:i) NO'>'emeer 2Q 12 PD HA3 ECL: Alert Initiating Canditian:
I NE! 99 0 I (Revision
Other conditions e>(ist 1tvhich in the judgment of the Emergency Director warrant declaration of an Alert. Operating Made Applieability:
: 6) NoYemeer 2012 Tab l e 9-FF-14: BWR-DAEC EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FAlALERT ANY Loss OR ANY Potential Loss of EITHER the Fuel Clad OR RCS barrierAey MY.Loss or ~any Potential Loss of either the Fuel Clad or OR RCS barrier. Fuel Cla d Barr i er L O SS POTENTIAL LOSS 1. RCS Activity A. Coolant activity Not Applicable greater than 300 !!Ci/gm dose eguivalent J-filA. (Site specific indications that reactor coolant activity is greater than 300 &#xb5;Ci/gm dose equivalent I +Mt FSl SITE AREA EMERGENCY Loss &rOR Potential Loss of ~ANY two barriers.
Not Applicable Example Emergeney Aetian L~1 els: (1) Other conditions e>(ist v,hich in the judgment of the Emergenc;'
RCS Barrier LOSS POTENTIAL LOSS 1. Primary Containment PressureConditions A. Primary Not Applicable containment pressure greater than (site specific &#xa5;a-tt:lej~
Director indicate that events are in progress or have occurred *which invohe an .actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. An;' releases are expected to be limited to small fractions of the EPA ProtectiYe Action Guideline exposure leYels. This IC addresses unanticipated conditions not addressed explicitly else'tvhere but that 1 Narrant declaration of an emergency because conditions e>(ist which are belie't 1 ed by the Emergency Director to fall under the emergency classification level description for an Alert. B-15}}
due to RCS leakage. 123 FGlGENERALEMERGENCY Loss of ANY two barriers AND Loss OR Potential Loss of the third barrierLoss of any MY_two barriers and Loss or QR.Potential Containment Barrier LOSS POTENTIAL LOSS 1. Primary Containment Conditions A. UNPLANNED A. Primary rapid drop in containmentTorus primary pressure greater containmentDrywe than (site specific ll pressure &#xa5;a-tt:lej 5 3 psi g following primary OR containmentDrywe B. Drywell or Torus 11 pressure rise H2 cannot be OR determined to be B. Primary LeSS +Ht\Nless containmentDrywe than 6% and 11 pressure Drywell &OR response not Torus 02 cannot consistent with be determined to LOCA conditions.
be less than 5% OR (site speeifie C. UNISOLABLE explosiYe direct downstream mixture) exists 12athway to the inside primary environment exists eontainment after 12rimary OR Fuel Clad Barrier RCS Barrier LOSS POTE NTIAL LOSS LOSS POTEN TIAL LOSS D. 124 NI;:I 99 0 I (Re*,*isioA 6) N01,cember 2012 Containment Barrier LOSS POTENTIAL LOSS containment C. HC+L (GraRh 4 of isolation signal EOP 2) exceeded.
OR Intentional Rrimary containment venting Rer EOPs I Fuel Clad Barrier RCS Barrier I LOSS I POTENTIAL LOSS LOSS POTENTIAL LOSS 2. RPV Water Level 2. RPV Water Level A. SAG entry is A. RPV water level A. RPV water level ot Applicable requiredPrimary cannot be restored cannot be restored eeRtaiRmeRt and maintained and maintained fleediRg required.
above Esite speeifie above fs-tte-RJH,l water le*,<el speeifie R::P\l eerrespeRdiRg te water le&#xa5;el the tep ef aetive eerrespeRdiRg te .fue.B+ 15 inches OR the tep ef aeti&#xa5;e cannot be fue-4+ 15 inches determined.
OR cannot be determined.  
: 3. Not Applicable
: 3. RCS Leak Rate Not Applicable Not Applicable A,_UNISOLABLE A. UNISOLABLE break in ,<\....~&#xa5; ef primary system the follewiRg:
leakage that Esite speeifie results in systems with exceeding the peteRtial fer high Max Normal eRergy liRe OQerating Limit breaks)Main (MNOL) of EOP Steam, HPCI, 3, Table 6 for Feedwater, EITHER of the RWCU, or RCIC following:
as indicated by the
* 1. Max failure of both :ti-formal isolation valves in OperatiRg aRyANY one line Temperature to close AND OR -EITHER:
* 2. Max
* High MSL flow or steam Nermal r\ * -A ---'-'r-.......... b --*----125 2. Jl-JEl 99 01 (ReYisioA  
: 6) Jl-JoYeA1ber 20 12 Containment Barrier LOSS POTENTIAL LOSS RPV Water Level Not Applicable A. SAG entry i s requiredPrimary eeRtaiRmeRt fleediRg required. 3. Primary Containment Isolation Failure A. UNISOLABLE Not Applicable Qrimary system leakage that results in exceeding the Max Safe OQerating Limit (MSOL) of EOP 3, Table 6 for EITHER of the following:
* TemQerature OR
* Radiation Level UNISO bABbe direet dewR s tream I Fuel Clad Barrier RCS Barrier I LOSS I POTENTIAL LOSS LOSS POTENTIAL LOSS tunnel Radiation tem12erature -be\.<el-Level
~ annunciators OR
* Direct re12ort of steam release OR -B. Emerge nc y RPV Depr ess urization required. l __ 126 NEJ 99 0 I (RevisioA
: 6) No,*emeer 20 12 Containment Barrier LOSS POTENTIAL LOSS pati:P,1t 1 ay te tile envirenment e*ists afteF pFimaF~' centainment iselatien signal OR ------B-:
Intentie nal 13FimaF~' centainment ventin g peF OR ----b7 l:l~HSQ LABLE pFimaFJ' system leakage that Fesl:llts in e~rneeEI in g El'.IIIER ef the fullevving:
L Ma* Safe QpeFating Tem13erntl:IFe. OR 2. Ma~E Safe Q13eFating AFea R:aEliatien I:,ev-eh I Fuel Clad Barrier RCS Barrier I LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 4. Primary Containment Radiation
: 4. Primary Containment Radiation A. Dr~ell Monitor Not Applicable A. Dr~ell Monitor Not Applicable (9I84A/B)
(9I84A/B) reading greater reading greater than 2000 R/hr. than 5 R/hr after OR reactor B. Torus Monitor shutdownk (9185A/B)
Primary reading greater ceAtaiAmeAt than 200 R/hr raEliatieA meAiter reaEliAg greater thaA (site specific value). 5. Other Indications
: 5. Other Indications A. Fuel damage Not A1;mlicableA,.
Not A1mlicableA,.
Not A1mlicableA,.
assessment (site specific as (site specific (site specific as indicates at least applicable) as applicable) applicable) 5% fuel clad damage.fstte-specific as applicable)
: 6. E m erge n cy Director Judgme n t 6. Emergency Director Judgment A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in the opinion of the the opinion of the the opinion of the the opinion of the Emergency Emergency E mergency Emergency Director that Director that Director that Dir ector that indicate s Loss of indicate s Potential indicates Loss of indicates Potential 127 4. NET 99 01 (Re,*isioR 6) No 1 ,ember ?Q 12 Containment Barrier LOSS POTENTIAL LOSS Primary Containment Radiation Not Applicable A. Dr~ell Monitor (9184A/B) reading greater than 5000 R/hr. OR B. Torus Monitor (9185A/B) reading greater than 500 R/hrA,. Primary ceAtaiAmeAt raEliatieA meAiter reaEliAg greater thaA (site specific 11*ah,1e). 5. Other Indications Not A1mlicableA,.
A. Fuel damage (site specific as assessment applicable)
~ASAP +.~j indicates at least 20% fuel clad damage.fstte-specific as applicable)
: 6. Emergency Director Judgment A. ANY condition in A. ANY condition in the opinion of the the opinion of the Emergency Emergency Director that Director that indicates Loss of indicates Potential Fuel Clad Barrier RCS Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS the Fuel Clad Loss of the Fuel the RCS Barrier. Loss of the RCS Barrier. Clad Barrier. Barrier. 128 NEJ 99 0 I (RevisioR
: 6) November 2012 Containment Barrier LOSS POTENTIAL LOSS the Containment Loss of the Barrier. Containment Barrier. 
 
Basis Information For BWR-DAEC EAL Fission Product Barrier Table 9-FF-1-i NET 99 0 I (Revision
: 6) 1'Jo,*ember 2012 BWR-DAEC FUEL CLAD BARRIER THRESHOLDS:
The Fuel Clad barrier consists of the zircalloy or stainless steel fuel bundle tubes that contain the fuel pellets. 1. RCS Activity Loss I.A This threshold indicates that RCS radioactivity concentration is greater than 300 &#xb5;Ci/gm dose equivalent J-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred , it represents a loss of the Fuel Clad Barrier. It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete.
Nonetheless, a sample-related threshold is included as a backup to other indications.
There is no Potential Loss threshold associated with RCS Activity.
Develeper Netes: Threshold values should be detennined assuming RCS radioactivity concentration equals 300 &#xb5;Ci/gm dose equivalent T 131. Other site specific units may be used (e.g., &#xb5;Ci/cc). Depending upon site specific capabilities , this threshold may have a sample analysis component and/or a radiation monitor reading component.
Add this paragraph (or similar wording) to the Basis if the threshold includes a sample analysis component , " lt is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete.
Nonetheless , a sample related threshold is included as a backup to other indications." 2. RPV Water Level 130 Loss 2.A The Loss threshold represents any EOP requirement for entry into the Severe Accident Guidelines.
NEI 99 01 (Re\*isioR e) No,*ember 201? This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured.The Loss threshold represents the EOP requirement for primary containment flooding.
This is identified in the BWROG EPGs/8AGs ,vhen the phrase , " Primary Containment flooding Is Required ," a-ppears.
Since a site specific RPV water level is not specified here , the Loss threshold phrase , " PrimaF)' containment flooding required ," also accommodates the EOP need to flood the primary containment v,hen RPV water level cannot be determined and core damage due to inadequate core cooling is believed to be occurring.
Potential Loss 2.A This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. 131 BWR FUEL CLAD BARRIER THRESHOLDS:
J!>JEI 99 o I (ReYisioR a) J!>Je,,eR1ber 2012 The RPV water level threshold is the same as RCS barrier Loss threshold 2.A. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.
132 DAEC FUEL CLAD BARRIER THRESHOLDS
{cont.): NEI 99 0 I (ReYisioR
: 6) }>fo,*ember 2012 This threshold is considered to be exceeded when , as specified in the site speeifie EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually , automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization.
EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed I imits. EOPs also specify depressurization of the RPV.in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of avai I able injection sources. Therefore , this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted , giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available , precluding RPV depressurization in an attempt to minimize loss of RPV inventory.
133
* I I I I I I DAEC FUEL CLAD BARRIER THRESHOLDS (eent.)::
Nel 99 O I (RevisioH a) }Jovember 2012 The term " cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.
In high-power A TWS/failure to scram events , EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier , the immediate need to reduce reactor power is the higher priority.
For such events , ICs ~SA6 or SS~&sect;_ will dictate the need for emergency classification.
Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier , a potential loss of the fuel clad barrier is specified. 134 BWR FUEL CLAD BARRIER THRESHOLDS:
DeYeloper Notes: Loss 2.A NEI 99 0 I (RevisioA
: 6) No 1 ,eA~eer 2012 The phrase , " Primary containment flooding required ," should be modified to agree with the site specific EOP phrase indicating e>(it from all EOPs and entry to the SAGs (e.g., dry,Nell flooding required , etc.). Potential Loss 2.A The decision that "RPV water level cannot be determined" is directed by guidance given in the RPV \vater level control sections of the EOPs. 3. Not Applicable (included for numbering consistency between barrier tables) 135 DAEC FUEL CLAD BARRIER THRESHOLDS
{cont.): 4. Primary Containment Radiation Loss 4.A and Loss 4.B Nel 99 O I (RevisioA
: 6) November 2012 The Drywell and Torus radiation monitor reading 2 correspond s to an instantaneous release of all reactor coolant mass into the Drywell or primary Toruscontainment , assuming that reactor coolant activity equals 300 &#xb5;Ci/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred , it represents a loss of the Fuel Clad Barrier. The radiation monitor readin~ in this threshold t&-are higher than that specified for RCS Barrier Loss threshold 4.A since it indicates a loss of both the Fuel C l ad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency. There is no Potential Loss threshold associated with Primary Containment Radiation. 136 B'WR DA.EC FUEL CLAD BARRIER THRESHOLDS (eoet.): NEI 99 () 1 (Re\*isioR
: 6) l>levember
?() 12 her Indications-..
: 5. 1. Other Iedieatioes Loss and/or Potential Loss 5.A Results obtained from procedure PASAP 7.2, Fuel Damage Assessment, indicate at least 5% fuel clad damage. ffhis subcategory addresses other site specific thresholds that may be included to indicate loss 2:!.:QQ!ential loss of the Fuel Clad barrier based o~la~ecific design characteristics not considered in the generic guidance.
There is no Potential Loss threshold associated with Other Indications.
Develof)er Notes: Loss and/or Potential Loss 5 .A Developers should determine if other reliable indicators exist to evaluate the status of this fission product barrier (e.g., reviev, accident analyses described in the site Final Safety Analysis Report, as updated).
The goal is to identify any unique or site specific indications that \Viii promote timely and accurate assessment of barrier status. Any added thresholds should represent appro>&#xa3;imately the same relative threat to the barrier as the other thresholds in this column. Basis infonnation for the other thresholds may be used to gauge the relative barrier threat level. &6_:...._Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost. Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentiall y lost in the event that barrier status cannot be monitored.
137 00 M ......
BWR--DAEC RCS BARRIER THRESHOLDS:
NEI 99 01 (RevisioR
: 6) No,*ember 2012 The RCS Barrier is the reactor coolant system pressure boundary and include s the RPV and all reactor coolant system piping up to and including the isolation valves. 1. Primary Containment PressureConditions Loss l.A The (site specific value) primary containment~
pressure is the drywell high pressure scram setpoint which indicates a LOCA by automatically initiating the-ECCS or equivalent makeup system . .:. There is no Potential Loss threshold associated with Primary Containment Pressure.
DeYel0per Notes: 2. RPV Water Level Loss 2.A This ,vater I+ 15 inches e;iel-corresponds to the top of active fuel (T AF) and is used in the E OPs to indicate challenge to core cooling. The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold 2.A. Thus , this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency.
This threshold is considered to be exceeded when , as specified in the site specific EOPs , RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually , automatically or by failure of the RCS barrier) or when procedural guidance or a lack of lo w pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a w ide choice of RPV injection sources to consider when restoring RPV water le vel to within prescribed limit s. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with lo w-pressure injection sources. In some events , elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore , this RCS barrier Loss is met only after either: 1) the RPV has been depres s urized , or required emergency RPV depressurization has been attempted , giving the operator an opportunity to assess the capability of l ow-pressure inject ion sources to restore RPV water level or 2) no low pressure RPV injection systems are available , precluding RPV depressurization in an attempt to minimize loss of RPV inventory.
139 B\&#xa5;R DAEC RCS BARRIER THRESHOLDS: "f)>ffiJ 99 0 I (RevisioA e) No 1 1emeer 2012 The term, " cannot be restored and maintained above ," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel , but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. 140 DAEC RCS BARRIER THRESHOLDS (cont.): NET 99 QI (Re,*ision 6) l>Joven~ser 2Q 12 In high-power ATWS/failure to scram events , EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCR WL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority.
For such events , ICs SAS or SSS will dictate the need for emer g ency classification.
There is no RCS Potential Loss threshold associated with RPV Water Level. 3. RCS Leak Rate Loss Threshold 3.A Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the retaining capability of the RCS until they are isolated.
If it is determined that the ruptured line cannot be promptly isolated from the Control Room , the RCS barrier Loss threshold is met. Loss Threshold 3.B Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed , the plant operators are directed to open safety relief valves (SRYs) and keep them open. Even though the RCS is being vented into the suppression pool , a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary. Potential Loss Threshold 3.A Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC , HPCI , etc., which indicate a direct path from the RCS to areas outside primary containment.
A Max Normal Operating Limit (MNOL) value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly.
141 L B\\'R DAEC RCS BARRIER THRESHOLDS:
NEI 99 0 I (Re,,*ision 6) November 2012 The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes , valves , and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the sy s tem. An UNISOLABLE leak which is indicated by Max Normal OperatingMNOL values escalates to a Site Area E mergency when combined with Containment Barrier Loss threshold 3.A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.
DAEC RCS BARRIER THRESHOLDS
{cont.): Developer Notes: Loss Threshold 3.A The list of systems included in this threshold should be the high energy lines which , if ruptured and remain unisolated , can rapidly depressurize the RPV. These lines are typically isolated by actuation of the Leak Detection system. Large high energy line breaks such as Main Steam Line (MSL), High Pressure Coolant Injection (HPCI), feedv,rater , Reactor Water Cleanup (RWCU), Isolation Condenser (JC) or Reactor Core Isolation Cooling (RCIC) that are UNTSOLABLE r e present a significant loss of the RCS barrier. 4. Primary Containment Radiation Loss 4.A The Drywellradiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment , assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 4.A since it indicates a loss of the RCS Barrier only. There is no Potential Loss threshold associated with Primary Containment Radiation. Developer Notes: The reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory , with RCS actiYity at Technical Specification allowable limits , into the primary containment atmosphere. Using RCS activity at Technical Spec i fication allowable limits aligns this threshold with TC SU3. Also , RCS activity at this level will typically result in primary containment 142 
 
B"WR RCS BARRIER THRESHOLDS:
NEl 99 01 (RevisioA
: 6) J>Jo 1 ,eA10er 2012 In some cases, the site specific physical location and sensitivity of the primary containment radiation monitor(s) may be such that radiation from a cloud of released RCS gases cannot be distinguished from radiation emanating from piping and components containing elevated reactor coolant activity. If so , refer to the Developer Guidance for Loss/Potential Loss 5.A and determine if an alternate indication is available.
: 5. Other Indications There are no Loss or Potential Loss thresholds associated with Other Indications.
DeYeloper Notes: Loss and/or Potential Loss 5 .A Developers should determine if other reliable indicators e>list to evaluate the status of this fission product barrier (e.g., reYiev,* accident analyses described in the site final Safety Analysis Report , as updated).
The goal is to identify any unique or site specific indications that 1tvill promote timely and accurate assessment of barrier status. Any added thresholds should represent apprmC:imately the same relative threat to the barrier as the other thresholds in this column. Basis information for the other thresholds may be used to gauge the relative barrier threat level. 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost. Potential Loss 6.A This thresho l d addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The E mergency Director should also consider whether or not to declare the barrier potentiall y lost in the event that barrier status cannot be monitored. Deyeloper Notes: 144 RWR DAEC CONTAINMENT BARRIER THRESHOLDS:
t'IBJ 99 O I (Re\*ision e) November 2012 The Primary Containment Barrier includes the drywell , the wetwell , their respective interconnecting paths , and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.
: 1. Primary Containment Conditions Loss l .A and l .B Rapid UNPLANNED loss of primary eontainment drywell pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of primary eontainmentdrywell integrity.
Primary containmentDrywell pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus , primary containmentdrywell pressure not increasing under these conditions indicates a Joss of primary containment integrity.
These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not a s signed. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Loss l.C The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). The existence of a filter is not considered in the threshold assessment.
Filters do not remove fission product noble gases. Jn addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiologica l releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category RI Cs. 145 DAEC CONTAINMENT BARRIER THRESHOLDS:
Loss l.D NEI 99 01 (Revision e) No*,ember 20 I 2 EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded.
Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.
Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment.
Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition.
DAEC CONTAINMENT BARRIER THRESHOLDS (cont.): DAEC CONTAINMENT BARRIER THRESHOLDS:
Potential Loss l .A The threshold pressure is the primary containmentTorus internal design pressure.
Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure.
A pressure of this magnitude is greater than those expected to result from any design basis accident and , thus , represent a Potential Loss of the Containment barrier. Potential Loss l .B If hydrogen concentration reaches or exceeds the lower flammability limit , as defined in plant EOPs , in an oxygen rich environment , a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment , loss of the Containment barrier could occur. Potential Loss l .C The Heat Capacity Temperature Limit (HC+L) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:
* Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized , OR 146 BWR CONTAINMENT BARRIER THRESHOLDS:
J!>J:El 99 Q 1 (Re., 1 isioA 6) J!>Jovemaer 2Q 12 -Suppression chamber pressure above Primary Containment Pressure Limit A , while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.
* The HC+L is a function of RPV pressure , suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore , the inability to maintain plant parameters below the limit constitutes a potential loss of containment.
147 DAEC CONTAINMENT BARRIER THRESHOLDS
{cont.):t DeYelof)er Notes: Potential Loss l .B NET 99 QI (Re*,ision
: 6) J!>l o,*emeer 2Q 12 BWR EPG s/SAGs s pecifically define the limit s a s sociated with e~&#xa3;plosive mixtures in terms ofdeflagration concentrations of hydrogen and oxygen. For Mk YIT containments the deflagration limits are "6% hydrogen and 5% m&#xa3;ygen in the drywell or suppression chamber". For Mk IJJ containments , the limit is the " Hydrogen Deflagration Overpressure Limit". The threshold term " explosive mixture" i s synonymous with the EPG/SAG " deflagration limits". Potential Loss l .C Since the HCTL is defined assuming a range of suppression pool water levels as low as the elevation of the downcomer openings in Mk J/11 containments , or 2 feet above the elevation of the horizontal vents in a Mk JII containment , it is unnecessary to consider separate Containment barrier Loss or :Potential Loss thresholds for abnormal suppression pool *Nater level conditions. If desired , developers may include a separate Containment Potential Loss threshold based on the inability to maintain suppression pool v,*ater leYel aboYe the downcomer openings in Mk I/Tl containments , or 2 feet above the ele*,ation of the horizontal vents in a Mk III containment
\Vith R..&deg;V pressure above the minimum decay heat removal pressure , if it will simplify the a s sessment of the suppression pool level component of the HCTL. 2. RPV Water Level There is no Loss threshold associated with RPV Water Level. Potential Loss 2.A The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold 2.A. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be restored and maintained and that core damage is possible.
BWR EPGs/SAGs specify the conditions that require primary containment flooding.
When primary containment flooding is required , the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to restqre and maintain adequate core cooling. 148 BWR CONTAINMENT BARRIER THRESHOLDS:
NEI 99 0 I (RevisioA
: 6) November 20 1 2 PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which , if not corrected , could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns , this threshold results in the declaration of a General Emergency.
DeyelepeF Netes: The phrase , " Primary coAtaiAmeAt floodiAg required ," should be modified to agree *.vith the site specific EOP phrase iAdicatiAg e>dt from all EOPs aAd eAtry to the SAGs (e.g., drywell floodiAg required , etc.). 3. Primary Containment Isolation Failure These thresholds address incomplete containment isolation that allows an UNJSOLABLE direct release to the environment.
Loss 3.A The use of the modifier " direct" iA defiAiAg the release path discrimiAates agaiAst release paths through iAterfaciAg liquid systems or miAor release pathways , such a s iAstrumeAt liAes , Aot protected by the Primary CoAtaiAmeAt JsolatioA System (PCIS). The e>ListeAce of a filter is Aot coAsidered iA the threshold assessmeAt.
Filters do Aot remove fissioA product Roble gases. IA additioA , a filter could become iAeffective due to iodiAe aAd/or particulate loadiAg beyoAd desigA limits (i.e., reteAtioA ability has beeA e>rneeded) or water saturatioA from steam/high humidity iA the release stream. FollowiAg the leakage of RCS mass iAto primary coAtaiAmeAt aAd a rise iA primary coAtaiAmeAt pressure , there may be miAor radiological releases associated
'Nith allovt1able primary coAtaiAmeAt leakage through various peAetratioAs or system compoAeAts. MiAor releases may also occur if a primary contaiAmeAt isolatioA valve(s) fails to close but the primar)' coAtaiAmeAt atmosphere escapes to aA eAclosed system. These releases do Aot coAstitute a loss or poteAtial loss of primary coAtaiAmeAt but should be evaluated usiAg the RecogAitioA Category A ICs. Loss 3.B EOPs may direct primary contaiAmeAt isolatioA valve logic(s) to be iAteAtioAally b)*passed , eveA if offsite radioactivity release rate limits will be eKceeded.
Under these coAditioAs with a valid primary coAtaiAmeAt isolatioA signal , the coAtaiAmeAt should also be coAsidered lost if primary coAtaiAmeAt veAtiAg is actually performed. JAteAtioAal ventiAg of primary coAtaimneAt for primary coAtaiAmeAt pressure or combustible gas coAtrol to the secoAdar)'
coAtaiAmeAt aAd/or the enviroAmeAt is a Loss of the CoAtaiAmeAt.
VeAtiAg for primary coAtaiAmeAt pressure coAtrol wheA Aot iA aA accideAt situatioA (e.g., to coAtrol pressure below the dr)"well high pressure scram setpoiAt) does Aot meet the threshold coAditioA.
149 Loss 3.GA NEr 99 QI (RevisioR a) 1l-Jo 1 , 1 ember 2012 The Max Safe Operating Limit (MSOL) for Temperature and the Ma>c. Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail , nor (2) personnel access necessary for the safe shutdown of the plant will be precluded.
EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.
BWR CONTA.INMENT BARRIER THRE8HOLD8:
The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes , valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. In combination with RCS potential loss 3.A this threshold would result in a Site Area Emergency.
There is no Potential Loss threshold associated with Primary Containment Isolation failureRCS Leak Rate. 150 DAEC CONTAINMENT BARRIER THRESHOLDS
{cont.):~
DeYeloper Notes: Loss 3.B NET 99 Q 1 (ReYisioA
: 6) ~Jo,,ember 2Q 12 Consideration may be given to speeifying the speeifie proeedural step within the Primary Containment Control EOP that defines intentional venting of the Primary Containment regardless of offsite radioaetivity release rate. 4. Primary Containment Radiation There is no Loss threshold associated with Primary Containment Radiation.
Potential Loss 4.A The dryw e ll radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary eontainmentdrywell , assuming that 20% of the fuel cladding has failed. The radiation monitor reading for the torus corresponds to an instantaneous release of all reactor coolant mass directly into the torus, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.
NUREG-1228 , Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents , indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring off site protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.
DeYeloper Notes: NUREG 1228 , Seurce Estimatiens Duringfncide,~t Respense te Se*rere Nuclear Pewer Pf.GmtAccidcnts , provides the basis for using the 20% fuel el adding failure value. Unless there is a site speeifie analysis justifying a different value , the reading should be determined assuming the instantaneous release and dispersal of the reaetor eoolant noble gas and iodine inventory assoeiated with 20% fuel elad failure into the primary eontainment atmosphere.
BWR CONTAINMENT Bz" .. RRIER THRESHOLDS:
: 5. Other Indications There is no Loss threshold associated with Other Indications Loss and/or Potential Loss 5.A 151 J!IJ:EI 99 () 1 (Re1,*isioA
: 6) J!llo 1 ,em0er 2012 Results obtained from procedure PASAP 7.2, Fuel Damage Assessment, indicate at least 25% fuel clad damage. This subeategory addresses other site specifie thresholds that may be iRcluded to iRdicate loss or poteRtial loss of the CoRtaiRmeRt barrier based OR plaRt specific desigR eharaeteristies Rot eoRsidered iR the geRerie guidaRee.
PASAP 7.2 only shows whether fuel damage is greater than or Jess than 25%, thus this indication is not likely to be declared before containment barrier potential loss 4.A which indicates 20% fuel damage. However, this potential loss threshold adds an additional layer of diversity to the scheme. 152 De,;eleper Notes: Loss and/or Potential Loss 5.A NEI 99 0 I (RevisioA
: 6) ~Jovember 2012 Developers should determine if other reliable indicators e1(ist to evaluate the status of this fission product barrier (e.g., review accident analyses described in the site Final Safety Analysis Report, as updated).
The goal is to identify any unique or site specific indications that will promote timely and accurate assessment of barrier status. Any added thresholds should represent approJ(imately the same relative threat to the barrier as the other thresholds in this column. Basis information for the other thresholds may be used to gauge the relafr,re barrier threat level. 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Containment barrier is lost. Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.
153 
 
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PWR CONTAINMENT BARRIER THRESHOLDS:
NEI 99 01 (ReYisioA
: 6) NoveA10er 2012 The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the ma i n steam , feedwater , and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the EGL from Alert to a Site Area Emergency or a General Emergency.
RCS or SC Tube Lealrnge Loss I .A This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment.
The condition of the SG , 1.vhether leaking or RUPTURED , is determined in accordance
\Vith the thresholds for RCS Barrier Potential Loss I .A and Loss I .A , respectively.
This condition represents a bypass of the containment barrier. FAULTED is a defined term within the NEI 99 01 methodology; this determination is not necessarily dependent upon entry into , or diagnostic steps within , an EOP. For example , if the pressure in a steam generator is decreasing uncontrollably fp*1rt &jthe FAULTED de.finitif:m]
and the faulted steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition , the steam generator is st il l considered FAULTED for emergency classification purposes. The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification.
Steam releases of this size are readily observable with normal Control Room indications. The lov,er bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU3 for the fuel clad barrier (i.e., RCS activity values) and IC SU4 for the RCS barrier (i.e., RCS leak rate values). This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant , or to drive an am,iliary' (emergency) feed 1.vater pump. These types of conditions 1.vill result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition).
The inability to isolate the steam flov, without an adverse effect on plant cooldovm meets the intent of a loss of containment.
Stearn releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period ohirne following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Stearn releases associated with the unexpected operation of a valve (e.g., a stuck open safety valve) do meet this threshold.
169 
 
PWR CONTAl}f}.4.ENT BAR.~..IER Thresholds:
lAadequate Heat Removal There is no Loss threshold assosiated with IRadequate Heat RemoYal. Potential Loss 2.A N~J 99 Ql (Re*,*ision 6) Jl,Joyenleer 2Q 12 This sondition represents an IMMINENT sore melt sequense whish , if not sorrested , sou Id lead to vessel failure and an insreased potential for oontainment failure. For this sondition to ossur, there must already have been a loss of the RC8 Barrier and the Fuel Clad Barrier. If implementation of a prosedure(s) to restore adequate sore sooling is not effestive (sussessful) within 15 minutes , it is assumed that the event trajestory will likely lead to sore melting and a subsequent shallenge of the Containment Barrier. The restoration prosedure is sonsidered "effestive" if sore exit thermosouple readings are desreasing and/or if reastor vessel level is insreasing. Whether or not the prosedure(s)
\Viii be effestive should be apparent within 15 minutes. The Emergensy Direstor should essalate the emergensy slassifisation level as soon as it is determined that the prosedure(s) will not be effestive.
Severe assident analyses (e.g., NUREG 1150) have sonoluded that funstion restoration prosedures san arrest sore degradation in a signifisant frastion of sore damage ssenarios , and that the likelihood of sontainment failure is \1 ery small in these events. Given this , it is appropriate to provide 15 minutes beyond the required entr)' point to determine if prosedural astions can reverse the sore melt sequence.
Developer Notes: 8ome site spesifis EOPs andtor EOP user guidelines may establish desision making criteria sonserning the number or other attributes of thermosouple readings nesessary to drive astions (e.g., 5 CETs reading greater than 1 , 200oF is required before transitioning to an inadequate sore cooling procedure).
To maintain consistency v,ith EOPs , these decision making sriteria may be used in the sore exit thermosouple reading thresholds.
Potential Loss 2.A. l Enter site specific sriteria requiring entry into a core sooling restoration prosedure or prompt implementation of core cooling restoration astions. A reading of 1 , 200oF on the CETs may also be used. for plants that have implemented Westinghouse Owners Group Emergency Response Guidelines , enter the parameters and values used in the Core Cooling Red Path. 172 (ll V'~ SSO'J SSOdAij JO , (l!.IB0lUf lU0WU!UlUO:)
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Figure 9 F 4: PWR Centainment Integrity er Byf-)ass Examf-)les Auxiliary Building I : : Effluent , I I Inside Containment Damper Damper Open valve Process Pump Cooling 178 Closed Cooling Water System NE! 99 0 I (Re,*ision 6) November 2012}}

Revision as of 13:44, 30 September 2018

Response to Request for Additional Information Regarding License Amendment Request (TSCR-166), Adoption of Emergency Action Level Scheme Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors. (Pa
ML18212A229
Person / Time
Site: Duane Arnold NextEra Energy icon.png
Issue date: 07/26/2018
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To:
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References
NG-18-0090
Download: ML18212A229 (187)
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Text

ATTACHMENT 1 NEXTERA ENERGY DUANE ARNOLD, LLC DUANE ARNOLD ENERGY CENTER RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATING TO LICENSE AMENDMENT REQUEST TSCR-166 UPDATED REDLINE MARKUP OF NEI 99-01 REVISION 6 311 pages follow NEI 99*0 1 [Revision

&] , lepment of Deve

  • Levels Emergency Action fer on-N Passive Reactors November 2012

[THIS PA.GE IS LEFT BL1i\NK INTENTIONA.LLY]

NEI 99 01 [Revision

&] Nuclear Energy Institute Duane Arnold Energy Center (DAEC) Emergency Action Levels Technical Bases Document TBD,2018 November 2012 Nue.'-ear Energy Institute , I 77G I Street N W , &.*ite 4()(), Washingten D. C. (2Q2. 739.8()()0)

ACKNOWLEDGMENTS This document 1 Nas prepared by the Nuclear Energy Institute (l'J:EI) Emergency Action Level (EAL) Task Foree. NEI Chairpersen:

David Young Preparatien Team Larry Baker E>celon Nuelear/Corporate Craig Banner PSEG Nuclear/Salem and Hope Creek Nuelear Generating Stations/US,A, John Egdorf Dominion Generation/Ke 1.vaunee Power Station Jade Lewis Entergy Nuelear/Corporate C. Kelly Walker Operations Support Ser11ices , Inc. Review Team Chris Boone Southern Nuelear/Corporate John Callahan Xcel Energ)s'Corporate/USA Bill Chausse Enereon Services , Inc. Kent Crocker Progress Energy/Brunswick

}foclear Plant Don Crowl Duke Energy/Corporate Roger Freeman Constellation Energy Nuclear Group/Corporate Walt Lee TVA Nuclear/Corporate Ken Meade FENOC/Corporate Don Mathena }l"e)ctEra Energy/Corporate David Stobaugh EP Consulting , LLC }tick Turner Cal !away Plant/STARS Maureen Za1,valick Diab lo Canyon Power Plant/STARS NOTICE Neither NEI , nor any of its employees , members , supporting organizations , contractors , or consultants make any warranty , expressed or implied , or assume any legal responsibility for the accuracy or completeness of , or assume any liability for damages resulting from any use of , any information apparatus , methods , or process disclosed in this report or that such may not infringe privately 01,vned rights. N u e l-ee1r Ener gy In st itute , J 77 6 J Slreet A'. W , Sti'ite 4{)(), W*1Shingten D. C. (2()2. 739.8()()Q)

EXECUTl\fE

SUMMARY

Jl>ffil 99 0 I (Re;*isioA 6) November 2012 Federal regulations require that a nuclear pov,1er plant operator develop a scheme for the classification of emergenC)' events and conditions.

This scheme is a fundamental component of an emergency plan in that it provides the defined thresholds that *will allow site personnel to rapidly implement a range of pre planned emergency response measures.

An emergency classification scheme also facilitates timely decision making by an Offsite Response Organization (ORO) concerning the implementation of precautionary or protective actions for the public. The purpose of Nuclear E nergy Institute (NEI) 99 01 is to provide guidance to nuclear power plant operators for the development of a site specific emergency classification scheme. The methodology described in this document is consistent with Federal regulations , and related US : Nuclear Regulatory Commission (NRG) requirements and guidance.

In particular , this methodology has been endorsed by the NRG as an acceptable approach to meeting the requirements of 10 CFR § 50.47(b)(4), related sections of 10 CFR § 50 , Appendix E , and the associated planning standard evaluation elements of}nJREG 0654/ FEMA REP 1 , Rev. 1 , Criteri*l for PrcpBr*ltien mui E?*l!u*ltien

&jRBmelegie8l Emergeney Respen s e P !*ins mui: Prcp8redness in Suppert f>jNuele*lr Pewer Pl*ll9ts , November 1980. NEI 99 01 contains a set of generic Initiating Conditions (ICs), Emergency Action Levels (EALs) and fission product barrier status thresholds.

It also includes supporting technical basis information , developer notes and recommended classification instructions for users. Users should implement ICs , EALs and thresholds that are as close as possible to the generic material presented in this document with allowance for changes necessary to address site specific considerations such as plant design , location , terminology , etc. Properly implemented , the guidance in JI.JEI 99 01 will yield a site specific emergency classification scheme 1 Nith clearly defined and r e adily observable EALs and thresholds.

Other benefits include the development of a sound basis document , the adoption of industry standard instructions for emergency classification (e.g., transient events , classification of multiple events , upgrading , downgrading , etc.), and incorporation of features to improve human performance.

An emergency classification using this scheme will be appropriate to the risk posed to plant workers and the public , and should be the same as that made by another NEI 99 01 user plant in response to a similar event. The individuals responsible for developing an emergency classification scheme are strongly encouraged to review all applicable NRG requirements and guidance prior to beginning their efforts. Questions concerning this document may be directed to the NEI Emergency Preparedness staff , NEI EAL task force members or submitted to the Emergency Preparedness Frequently Asked Questions process. Finally , unique State and local requirements associated with an emergency classification scheme are not reflected in this guidance.

Incorporation of these requirements may be performed on a case by case basis in conjunction with the appropriate ORO agency. Any such changes will require a review under the applicable sections of 10 CFR 50.

},JEJ 99 g 1 (Re¥isioA a) },foyember 2912 [THIS PAGE IS LEFT BLANK INTENTIONA.LLY]

ll TABLE OF CONTENTS NEI 99 0 l (Re*,*isioA e) l>Jo>,*emeer 2012 EXECUTIJJE

SUMMARY

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

i 1 BASIS FOR EMERGENCY ACTION LEVELS .................................................................

1 1.1 OPERATING REACTORS ****.******....**.......*...*.*.......*****..*.***....****.***.***..*.***...*.*.*.*....****.******

1 1.2 INDEPENDENT SPENT FUEL STORAGE INSTALLATION

{ISFSI) .....................................

2 1.3 NRC ORDER EA-12-051

4 2 KEY TERMINOLOGY USED IN DAEC EAL SCHEME .....................................................

6 2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) ...............................................................

6 2.2 INITIATING CONDITION (IC) ..........................................................................................

8 2.3 EMERGENCY ACTION LEVEL (EAL) ...............

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

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

8 2.4 FISSION PRODUCT BARRIER THRESHOLD

  • ...*****.*....*..****.*.....**.*..***.*..***..***.*******..*.*.**.**

8 3 DESIGN OF THE DAEC EMERGENCY CLASSIFICATION SCHEME ...........................

11 3 .1 ASSIGNMENT OF EMERGENCY CLASSIFICATION LEVELS (ECLS) .............................

11 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS **.*..**************

17 3.3 DAEC-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION 18 3.4 IC AND EAL MODE APPLICABILITY

                • .*.*....*****.*..***.....****...***...***.****....****..*........**.

20 4 DAEC SCHEME DEVELOPMENT

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

22 4.1 GENERAL DEVELOPMENT PROCESS ***************************************

                                                    • .**********

22 4.2 CRITICAL CHARACTERISTICS

..**********.*.......*****.*.***...*..***....***...****...*..***.*...*.**************.

23 4.3 INSTRUMENTATION USED FOR EALs ******************..************...***.*.*****.*.*****.*.*....************

25 4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA

                        • ..

27 5 GUIDANCE ON USING THE DAEC EALS *........*.....*..*....*............................................

29 5 .1 GENERAL CONSIDERATIONS

                                                      • .************************************************************

29 5.2 CLASSIFICATION METHODOLOGY

            • .****..***.*********....****....****..****....***......***.....**....***.

31 5.3 CLASSIFICATION OF MULTIPLE EVENTS AND CONDITJONS

    • ...****...***...***...**......********

31 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION

  • .*.**.****.**.******.*.******

32 5.5 CLASSIFICATION OF IMMINENT CONDITIONS

...*****.**..*.*.**.****.**...****..****..*.******........***

33 5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING

      • ..............

33 5.7 CLASSIFICATION OF SHORT-LIVED EVENTS *****************************.*******.******************..*.***

35 5.8 CLASSIFICATION OF TRANSIENT CONDITIONS

  • ......*****.....***...*.***..**....****...*****************

35 5.9 AFTER-THE-FACT DISCOVERY OF AN EMERGENCY EVENT OR CONDITJON

    • ...*.*..*...

36 lll "NEI 99 01 (Re*,isieR e) tfo't'emeer 2012 5.10 RETRACTION OF AN EMERGENCYDECLARATION

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

36 6 ABNORMAL RAD LEVELS / RADIOLOGICAL EFFLUENT ICS/EALS ........................

37 7 COLD SHUTDOWN / REFUELING SYSTEM MALFUNCTION ICS/EALS ....................

74 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION

{ISFSI) ICS/EALS ............

117 9 FISSION PRODUCT BARRIER ICS EALS ................................................................

120 10 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS ....... 179 11 SYSTEM MALFUNCTION ICS EALS *************************************************************************

215 APPENDIX A -ACRONYMS AND ABBREVIATIONS

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

A*1 APPENDIX B -DEFINITIONS

B-1 I V NEI 99 0 I (ReYisioA e) NoYember ?Q 12 DEVELOMENT OF DUANE ARNOLD EMERGENCY ACTION LEVELS FOR NON PASSIVE REACTORS TECHNICAL BASIS DOCUMENT 1 BASIS FOR EMERGENCY ACTION LEVELSREGULATORY BACKGROUND 1.1 OPERA TING REACTORS Title 10, Code of Fe d era l Regul a tions (CFR), E nergy , contains th e U.S. Nuc l ear Re g ulatory Commission (NRC) regulation s that apply to nuclear power fac iliti es. Several of th ese regulations gove rn various aspects of an emergency class i fication scheme. A revi ew of the relevant sect i ons li ste d below wi ll aid the reader in under s tandin g the key terminolo gy provided in Sect i on 3.0 of this do c ument.

  • 10 CF R § 50.47(a)(l)(i)
  • 10 CFR § 50.47(b)(4)
  • IO CFR § 50.54(q)
  • 10 CF R § 50.72(a)
  • 10 CF R § 50, Append i x E, JV.B , Assessment Actions
  • 10 CFR § 50, Append i x E, IV.C, Act i vat i on of Emergency Organization The above re g ul at ions are s uppl emented by var iou s regulatory guidance documents. Three documents of particular relevance to NEI 99-01 are: NUREG-0654/FEMA

-REP-l, Criteria fo r Preparation and Evaluation of R adio logi cal Emergency R esponse Plans and Pr eparedness in Support of Nuclear Power Plants, October 1980. [Refer to Appendix 1, Emergency Action Level Guidelines for Nu cl ea r Po wer Plant s] NUREG-1022 , Eve nt R eporting Guidelines JO CFR § 50. 7 2 and§ 50. 7 3 Re g ulator y Guide 1.101 , Eme r ge n cy Response Planning a nd Preparedness for Nuclear Power R eact or s 1 1.2 NEI 99 01 (Re*,*isioR e) November 2012 The above list is not all inclusive and it is strongly recommended that scheme developers consult with licensing/regulatory compliance personnel to identify and understand all applicable requirements and guidance. Questions may also be directed to the NEI Emergency Preparedness staff. PERMANENTLY DEFUELED STATION NEI 99 01 provides guidance for an emergency classification scheme applicable to a permanently defueled station. This is a station that generated spent fuel under a 10 CFR § 50 license, has permanently ceased operations and will store the spent fuel onsite for an eJ(tended period of time. The emergency classification levels applicable to this type of station are consistent

'Nith the requirements of 10 CFR § 50 and the guidance in NUREG 0654/FEMA REP 1. In order to relaJ( the emergency plan requirements applicable to an operating station , the O'.vner of a permanently defueled station must demonstrate that no credible event can result in a significant radiological release beyond the site boundary.

It is e>(pected that this verification 1,vill confirm that the source term and motive force available in the permanently defueled condition are insufficient to warrant classifications of a Site Area Emergency or General Emergency.

Therefore , the generic Initiating Conditions (ICs) and Emergency Action Levels (EALs) applicable to a permanently defueled station may result in either a Notification of Unusual EYent (NOUE) or an Alert classification.

The generic ICs and EALs are presented in Appendi1( C , Perm6lnently Defueled St6ltion K J s/EALs. e.1.L 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 f-50 and the guidance in NUREG 0654/FEMA-REP-l.

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 f-50.47 emergency plan. The generic ICs and EALs for an ISFSI are presented in Section 8 , ISFSI ICs/EALs. IC E-HUl covers the spectrum of credible natural and man-made events included within the scope of an ISFSI design. This IC is not applicable to installations or facilities that may process and/or repackage spent fuel (e.g., a Monitored Retrievable Storage Facility or an ISFSI at a spent fuel processing facility).

In addition, appropriate aspects oflC HUI and IC HAI should also be included to address a HOSTILE ACTION directed against an ISFSI. The analysis of potential onsite and offsite consequences of accidental releases associated with the operation of an ISFSI is contained in NUREG-1140, A Regulatory Analysis on Emergency Preparedness for Fuel Cycle and Other Radioactive Material Licensees.

NUREG-1140 concluded that the postulated worst-case accident involving an ISFSI has insignificant consequences to public health and safety. This evaluation shows that the 2 j l>TEI 99 0 I (RevisioR a) November 2012 maximum offsite dose to a member of the public due to an acc id ental r e l ease of radioact i ve materials would not exceed 1 rem Effective Dose Equivalent.

Regard in g the above in formation, the expectat i o n s for an offsite response to an A l ert classified under a 10 CFR f-72.32 emergency plan are genera ll y consistent w ith those for a Notification of Unusual Event in a 10 CFR f-50.47 emergency plan (e.g., to provide assistance ifrequested).

Also, the li censee's Emergency Response Organization (ERO) required for 10 CFR f-72.32 emergency plan is different than that prescribed for a 10 CFR f-50.47 emergency plan (e.g., no emergency technical support function).

3

+:4l]_NRC ORDER EA-12-051 NEI 99 QI (Re,*isio A e) }fo,*emaer 2()12 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 functions of core cooling and containment integrity , and ultimately led to core damage in three reactor s. 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 severa l 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, Issuanc e of Order to Modify Licenses with R eg ard to Reliabl e Spent Fuel Pool In strumentation, on March 12 , 2012 , to all US nuclear plants with an operating license , construction permit , or combined construction and operating license. NRC Order EA-12-051 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 pro vide 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;
  • 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 , Indu stry Guidance for Compliance with NRC Order EA-12-051, " To Modify Licenses with Regard to Reliabl e Spent Fuel Pool In strume ntation , " provides guidance for complying with NRC Order EA-12-051. NEI 99-01, Revi sio n 6 , include s three EALs that reflect the availability of the enhanced spent fuel pool level instrumentation associated with NRC Order EA-12-051. These EALs are included within existing IC§. AA+/-RA2 , and new ICs AS2 RS2, and AmRG2. Associated EAL notes , bases and developer notes are also provided. It is recommended that these EALs be implemented when the enhanced spent fuel pool level instrumentation is available for use. 4 I Tke regelater

' >HS I 99 01 (Rev;,;,.

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"'ith th' , ' ,a '" aeeer<laaee with IQ CF&* aage ersabffiit itte tRe NRG 1.S---,:L:nP:L:l:C:A:B:IL:l:T:Y~T:O~" ~"ANEEl>-ANllsSl""'""M<~:::~

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  • L r e plaa
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~.~'., ~~m,aaeatly SefuoloS , as ef 2~~12a(uclear power reactors in

  • ea referreS 10 as J"' ' "ever , 11 may be a<la teS I se ealleS J -"' :'**.sifieatiea seReme fe~:*"era:,'

."" plaat Sesigas) :, wel~ 9tiv~~~eS

""" passive Sesiges ev1at10as from tRe geaerie a :aaeeS ROA passive reaet~r o, o epors efaa effiergeao

' and criteria , and operating

~a,Saaee te aeeeaR! fer tRo a* ti'.'""' may aeeS te propose ; , ,ties , behNeen t"a fl1 me ers --ane--J generation plants. c aractenst1cs and capab T . I ~rences m design13ara t There are significant d . power plants (of an , es1gn ~nd operating differences ia searoe tefffi). Fe~ :~:::~:::~

~: !mall MeSalar :.:::~"(~a~e~mere_ial Raelea, ocument is not applicabl t S e.g., differences e o MRs. 5 NE! 99 01 (RevisioA
6) }Joyemeer 2012 2 KEY TERMINOLOGY USED IN NEI 99 01DAEC EAL SCHEME There are severa l key terms that appear throughout t h e NEI 99 OIEAL methodology.

These terms are introduced in th i s sect i on to support und ersta ndin g of s ub seque nt material.

As an a id to the reader , the following table is provided as an overview to illustrate the relat i onship of the terms to each ot h er. Emergency C l assificatio n Leve l Unusua l Event I Alert I SAE I GE In i t i at in g Cond iti o n In iti at in g Cond i tio n Initiati n g Cond i t i on Initiating Co ndi t i o n Emergency Act i on Emergency Actio n Emergency Action Emergency Action Leve l (1) Leve l (1) Leve l (1) Level (1)

  • Operating Mode
  • Operating Mode
  • Operati n g Mode
  • Operating Mode App li cability App li cability App li cability App li cabi li ty
  • Notes
  • Notes
  • Notes
  • Notes
  • Basis
  • Basis
  • Basis
  • Basis (1) -When making an emerge n cy classification, the Eme r gency Director must consider a ll information having a bearing o n the proper assessment of an Initiating Condition. T hi s includes the Emergency Act i on Leve l (EAL) plus the associated Operating Mode Applica bilit y, Notes and th e in formi n g Basis infor m ation. In the Recognition Category F matrices, EALs are referred to as Fission Prod u ct Barrier T h resholds; the thresho ld s serve the same function as a n EAL ,-., 2.1 EMERGENCY CLASSIFICATION LEVEL (ECL) One of a set of names or titles estab li shed by the US Nuclear Regulatory Co mmi ssion (NRC) for grouping off-norma l events or condit i o n s accordi n g to (1) potential or act u a l effects or consequences , and (2) resulting ons i te a nd offsite response act i ons. The emergency classification l evels , in ascending order of severity, are: Notificat i on of Un u sua l Event (NOUE) Alert Site Area Emerge n cy (SAE) Genera l Emergency (GE) I 2.1.1 Notification of Unusua l Event (NOUE)4 Events are in progress or h ave occ urr ed whic h indicate a potential degradation of the level of safety of the p l ant or indicate a security threat to faci lit y protection has been in itiated. No releases of radioactive m aterial requiring offsite response or m o nit oring are expected unl ess furt h er de gradat i o n of safety systemsSAFETY SYSTEMS occurs. This term is sometimes shorteAeEI to Ummial E,*eAt (UE) or other similar site s13eeifie termiAology.

The terms }JotifieatioA ofUAusual E\*eAt , NOUE anel Unu s ual EYeAt are u s eel iAterehaAgeaely throughout this eloeumeAt 6

NEl 99 0 I (ReYision

6) ~lo .. *emeer 2012 Purpose: The purpose of this classification is to assure that the first step in future response has been carried out , to bring the operations staff to a state of readiness, and to provide systematic handling of unusual event information and decision-making.

2.1.2 Alert Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels. Purpose: The purpose of this classification is to assure that emergency personnel are readily available to respond if the situation becomes more serious or to perform confirmatory radiation monitoring ifrequired , and provide offsite authorities current information on plant status and parameters.

2.1.3 Site Area Emergency Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts; 1) toward site personnel or equipment that could lead to the likely failure of or; 2) that prevent effective access to , equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PAG exposure levels beyond the site boundary.

Purpose: The purpose of the Site Area Emergency declaration is to assure that emergency response centers are staffed , to assure that monitoring teams are dispatched , to assure that personnel required for evacuation of near-site areas are at duty stations if the situation becomes more serious , to provide consultation with offsite authorities , and to provide updates to the public through government authorities.

2.1.4 General Emergency (GE) Events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility.

Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area. Purpose: The purpose of the General Emergency declaration is to initiate predetermined protective actions for the public , to provide continuous assessment of information from the licensee and offsite organizational measurements, to initiate additional measures as indicated by actual or potential releases , to provide consultation with offsite authorities, and to provide updates for the public through government authorities. 7 2.2 INITIATING CONDITION (IC) NE! 99 01 (R1:wisioA e) November 2012 An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.

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 as a continuous , measurable parameter (e.g., RCS leakage), an event (e.g., an earthquake) or the status of one or more fission product barriers (e.g., loss of the RCS barrier).

Appendix 1 of NUREG-0654 does not contain example Emergency Action Levels_ (EALs) for each ECL , but rather Initiating Conditions (i.e., plant conditions that indicate that a radiological emergency , or events that could lead to a radiological emergency, has occurred).

NUREG-0654 states that the Initiating Conditions form the basis for establishment by a licensee of the specific plant instrumentation readings (as applicable) which , if exceeded , would initiate the emergency classification. Thus , it is the specific instrument readings that would be the EALs. Considerations for the assignment of a particular Initiating Condition to an emergency classification level are discussed in Section 3. 2.3 EMERGENCY ACTION LEVEL (EAL) A pre-determined, site-specific, observable threshold for an Initiating Condition that , when met or exceeded, places the plant in a given emergency classification level. Discussion:

EAL statements may utilize a variety of criteria including instrument readings and status indications; observable events; results of calculations and analyses; entry into particular procedures; and the occurrence of natural phenomena.

2.4 FISSION PRODUCT BARRIER THRESHOLD A pre-determined , site-specific , observable threshold indicating the loss or potential loss of a fission product barrier. Discussion:

Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment.

This concept relies on multiple physical barriers , any one of which , if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.

The primary fission product barriers are: Fuel Clad Reactor Coolant System (RCS) Containment

---Upon determination that one or more fission product barrier thresholds have been exceeded , the combination of barrier loss and/or potential loss thresholds is compared to the fission product barrier IC/EAL criteria to determine the appropriate ECL. In some accident sequences, the ICs and EALs presented in the Abnormal Radiation Levels/ Radiological Effluent (AR) Recognition Category will be exceeded at the same 8 NEI 99 QI (Re~*isioR e) No~*ember 2Q 12 time , or short l y after, the loss of one or more fission product barriers.

T hi s redundancy is intentional as the former ICs address radioactiv it y releases that result in certain offs it e doses from whatever ca u se, including events that might not be fully encompassed by fission product barriers (e.g., spe nt fuel pool acc id ents , design containment l eakage following a LOCA , etc.). 9 10 NEI 99 QI (Re\*isioR e) }>Joyemaer 2Ql2

}ffil 99 0 I (Revision

6) }!oveFRber 2012 3 DESIGN OF THE NEI 88 01 DAEC EMERGENCY CLASSIFICATION SCHEME 3.1 ASSIGNMENT OF EME R GENCY CLASS I FICATION LEVELS (ECLS) An effective emergency c l ass ifi cat i o n sc h eme mu st in corporate a realistic a nd accurate assessment of r i sk , both to plant wo rk e r s and the public. T h ere are obvio u s h ealth a nd safety risks in underestimating the potential or act ual threat from an event or cond i tio n; however , there are a l so ri sks in overest im at in g t h e threat as we ll (e.g., harm that may occur during an evac u ation). T h e 1'IBI 99 OlDAEC emerge n cy classification sc h eme attempts to strike an appropriate balance between reasonably anticipated event or conditio n consequences , potential acc id ent trajector i es , and risk avoida n ce or minimization. There are a range of " n o n-emergency events" reported to the US N ucl ear Regulatory Commiss i o n (NRC) staff in accordance wit h the requirements of 10 CFR f-50.72. Guidance concerning these reporting requirements , and exa mpl e events , are provided in NUREG-1022. Certain events reportable under the provisions of 10 CFR f-50.72 may a l so require the declaration of a n eme r gency. In order to a li gn each Initiating Conditions (IC) with the appropriate ECL , it was n ecessary to determine the attr ibut es of each ECL. The goa l of this process i s to answer the quest i on , " What events or cond iti ons shou ld be placed under each ECL ?" The follow in g sources provided in formation and context for the development of ECL attributes.

Assessments of the effects a nd consequences of different t ypes of eve nt s a nd cond iti ons Typical DAEC abnorma l and e m ergency operating procedure setpoi nt s an d transition cr it er i a Typical DAEC Technical Specification limits and contro l s Radiological Effluent Tedrnical Specifications (RET8)/0ffsite Dose Caleulation Assessment Man u a l (ODAM) radiological release limits Rev i ew of se l ecte d Updated F inal Safety A n a l ysis Report (UFSAR) accident ana l yses E nvironmental Protection Age n cy (E PA) Protective Actio n Guidelines (PA Gs) NUREG 0654 , Appendi x 1 , Emerg e nc y Action Lev e l Guide l ine s for Nuclear Power Plants Industry Operating Experience Input from industry DAEC s ubj ect matter experts and NRG staff members T h e fo ll owi n g ECL attrib ut es were-are created used by the Revision 6 Preparation Team to aid in the development oflCs a nd Emerge n cy Act i on Leve l s (EALs). The team deeided to inelude the attributes in this revision since theyThe attributes may be useful in bri efing and training setti n gs (e.g., h e lpin g an E m ergency Director understand why a particular condit i on i s classified as an A l ert). It should be stressed that developers not attempt to redefine these attributes or apply them in any fashion that would change the generic guidance contained in this document\

+ The use of EGL attributes is at the eliseretion ofa lieensee anel is not a requireFRent of the J!-lRC. Ifa lieensee ehooses in ineor13orate the EGL attributes into their seheFRe basis eloeuFRent , it t'l'lust be very elear that the NRG staff has not enelorseel their aeee13tability or a1313lieation for any 13ur13ose.

In 13artieular , the staff eloes not eonsieler the 11 tlEI 99 01 (Re11isioA

6) }>fo*reFAber 20] 2 attribute stateFAeAts to superse ee es justifyiAg EAL ohaAges is uAaooeptable.

As a result , the use of the attributes as a basis for a th tablishea EGL aefiAitiOAS

.. 12 The attributes of each EGL are preseAted below. 13 ~IEI 99 QI (Revision

6) ~foyemaer 2Q12 3.1.1 Notification of Unusua l Event (NOUE) l>ffil 99 0 I (Re*,*isioR e) l>Jo1rember 2012 A Notificat i on of Un u sua l Event , as defined in section 2.1.1, in c lud es but is not limit ed to an event or condition that in volves: (A) A precursor to a more significant event or condition. (B) A minor loss of control of rad i oactive materials or the ability to contro l radiation levels w ithin the plant. (C) A consequence otherwise significant enoug h to warrant notification to l ocal , State and Federa l authorities. 3.1.2 Alert An Alert, as defined in sect i on 2.1.2, inc lud es but i s not limit ed to an event or condition that involves: (A) A loss or potential lo ss of e i ther the fuel clad or Reactor Coo l ant System (RCS) fissio n product barrier. (B) An event or condition that significantly reduces the margin to a loss or potential l oss of the fuel c l ad or RCS fission produ ct barrier. (C) A sign ifi cant loss of control ofrad i oactive materials resulting in an in abi lit y to contro l radiation l eve l s w i th in the plant , or a release of radioactive materials to the environment that cou ld result in d oses greater than I% of an EPA PAG at or beyond the s it e boundary. (D)A HOSTILE ACTION occurr in g w ithin the OWNER CONTROLLED AREA , including those directed at an Independent Spent Fuel Storage In s tallat i on (ISFSI). I 3.1.3 Site Area Emergency (SAE) A Site Area Emergency , as defined in section 2.1.3 , includes but is not limited to an event or cond ition that involves: (A)A l oss or potential l oss of any two fission product barriers -fue l clad , RCS and/or contai nm ent. (B) A precursor event o r condit i on that may l ead to the loss or potential lo ss of : multiple fission product barriers wit hin a relatively short period of time. Prec ur sor events and cond ition s of this type in c lud e those that c hall enge the monitoring and/or contro l of multip l e safety systemsSAFETY SYSTEMS. (C) A release of radioact i ve materials to the env ironm ent t h at could result in doses greater than 10% of an EPA PAG at or beyond the site boundary. (D)A HOSTILE ACTION occurr in g w ithin the plant PROTECTED AREA. 14

~lei 99 g 1 (Re~*isioR e) tJoyemaer 2912 3.1.4 General Emergency (GE) I 3.1.5 A General Emergency, as defined in section 2.1.4, includes but is not limited to an event or condition that involv es: (A)Loss of any two fission product barrier s AND lo ss or potential lo ss of the third barrier -fuel clad, RCS and/or containment. (B) A precursor event or condition that , unmitigated , may lead to a loss of all three fission product barrier s. Precur so r events and conditions of this type include tho se that lead directly to core damage a nd lo ss of containment int egr ity. (C) A release of radioactive materials to the environment that could result in doses greater than an EPA PA G at or beyond the site boundary. (D) A HOSTIL E ACTION resulting in the loss of key safety functions (reactivity control , core cooling/RPV water l eve l or RCS heat removal) or damage to spent fuel. 15 Risk-Informed Insights NEI 99 0 l (Re,*isioA

6) No11ember 2012 Emergency preparedness is a defense-in-depth measure that is independent of the assessed risk from any particular accident sequence; however , the development of an effective emergency classification scheme can benefit from a review of risk-based assessment results. To that end, the development and assignment of certain I Cs and EALs also considered insights from several site-specific probabilistic safety assessments (PSA also lmovm as probabilistic risk assessment, PR.A,). Some generic insights from this review included:
1. Accident sequences involving a prolonged loss of all AC power are significant contributors to core damage frequency at many Pressurized Water Reactors (PWRs) aHe-Boiling Water Reactors (BWRs). For this reason, a loss of all AC power for greater than 15 minutes , with the plant at or above Hot Shutdown , was assigned an ECL of Site Area Emergency.

Precursor events to a loss of all AC power were also included as an Unusual Event and an Alert. A station blackout coping analyses performed in response to 10 CFR f-50.63 and Regulatory Guide 1.155, Station Blackout , may be used to determine a time-based criterion to demarcate between a Site Area Emergency and a General Emergency.

The time dimension is critical to a properly anticipatory emergency declaration since the goal is to maximize the time available for State and local officials to develop and implement offsite protective actions. 2. For severe core damage events , uncertainties exist in phenomena important to accident progressions leading to containment failure. _Because of these uncertainties , predicting the status of containment integrity may be difficult under severe accident conditions.

_This is why maintaining containment integrity alone following sequences leading to severe core damage is an insufficient basis for not escalating to a General Emergency.

3. PSAs indicated that leading contributors to latent fatalities were sequences involving a containment bypass, a large Loss of Coolant Accident (LOCA) with early containment failure , a Station Blackout lasting longer than the site specificDAEC coping period , and a reactor coolant pump seal failure. The generic EAL methodology needs to be sufficiently rigorous to address these sequences in a timely fashion. 16 tiEI 99 O I (Re\*isioR e) N0\'0ffi80F 2012 3.2 TYPES OF INITIATING CONDITIONS AND EMERGENCY ACTION LEVELS The NEI 99-01 methodology makes use of symptom-based , barrier-based and based ICs and EALs. Each type is discussed below. Symptom-based ICs and EALs are parameters or conditions that are measurable over some range using plant instrumentation (e.g., core temperature , reactor coolant level , radiological effluent, etc.). When one or more of these parameters or conditions are normal , reactor operators will implement procedures to identify the probable cause(s) and take corrective action. Fission product barrier-based I Cs and EALs are the subset of symptom-based EALs that refer specifically to the level of challenge to the principal barriers against the release of radioactive material from the reactor core to the environment.

These barriers are the fuel cladding , the reactor coolant system pressure boundary , and the containment.

The barrier:-based I Cs and EALs consider the level of challenge to each individual barrier -potentially lost and lost -and the total number of barriers under challenge.

Event-based I Cs and EA Ls define a variety of specific occurrences that have potential or actual safety significance.

These include the failure of an automatic reactor scram ftftj3 to shut down the reactor , natural phenomena (e.g., an earthquake), or man-made hazards such as a toxic gas release. 17 3.3 N888 DESIGN DIFFERENCES l'ffil 99 Q 1 (Re,*isioR

6) NoYemeer 2QJ2 The NEI 99 01 emergency classification scheme accounts for the design differences bet\veen PWRs and BWRs by specifying EALs unique to each type of Nuclear Steam Supply System (NSSS). There are also significant design differences among PWR NSSSs; therefore , guidance is provided to aid in the development ofEALs appropriate to different PWR NSSS types. Where necessary , development guidance also addresses unique considerations for advanced non passive reactor designs such as the Advanced Boiling Water Reactor (ABWR), the Advanced Pressurized Water Reactor (AP\VR) and the Evolutionary Power Reactor (EPR). Developers 1 Nill need to consider the relevant aspects of their plant's design and operating characteristics when converting the generic guidance of this document into a site specific classification scheme. The goal is to maintain as much fidelity as possible to the intent of generic ICs and EALs within the constraints imposed by the plant design and operating characteristics. To this end , developers of a scheme for an advanced non passive reactor may need to add , modify or delete some information contained in this document; these changes will be reviewed for acceptability by the NRG as part of the scheme approval process. The guidance in NEI 99 01 is not applicable to advanced passive light water reactor designs. An Emergency Classification Scheme for this type of plant should be developed in accordance with NEI 07 01 , },/ethedetogyfer Develepment of Emergency Actien Le*,iels , Ad 1;cmced Pc1ssi>;e Light Water Rec1cters.

J.:..43.3 DAEC-SPECIFIC ORGANIZATION AND PRESENTATION OF GENERIC INFORMATION The scheme's generic information is organized by Recognition Category in the following order. A-R -Abnormal Radiation Levels/ Radiological Effluent -Section 6 C -Cold Shutdown/

Refueling System Malfunction

-Section 7 E -Independent Spent Fuel Storage Installation (ISFSI) -Section 8 F -Fission Product Barrier -Section 9 H -Hazards and Other Conditions Affecting Plant Safety-Section 10 S -System Malfunction

-Section 11 PD Permanently Defueled Station Appendi><

C Each Recognition Category section contains a matrix showing the ICs and their associated emergency classification levels. The following information and guidance is provided for each IC: ECL -the assigned emergency classification level for the IC. Initiating Condition

-provides a summary description of the emergency event or condition.

Operating Mode Applicability

-Lists the modes during which the IC and associated EAL(s) are applicable (i.e., are to be used to classify events or conditions).

18 tlEI 99 QI (RevisioR e) tlo\'efl'!eer 2Ql2 Example Emergency Action Level(s) -Provides examples of reports and indications that are considered to meet the intent of the IC. Developers sliould address eacli e~(ample EAL. If tlie generic approacli to tlie development of an e>i,aFHple EAL cannot be used (e.g., an assumed instrumentation range is not available at tlie plant), tlie developer sliould atteFHpt to specify an alternate means for identifying entry into tlie IC. For Recognition Category F, the fission product barrier thresholds are presented in tables applicable to BWRs and PWRs, and arranged by fission product barrier and the degree of barrier challenge (i.e.,_-potential loss or loss). This presentation method shows the synergism among the thresholds , and supports accurate assessments.

Basis -Provides background information that explains the intent and application of the IC and EALs. In some cases , the basis also includes relevant source information and references. 19 NEI 99 01 (ReyisioA

6) }!o 1 ,emeer 2012 Developer Notes Information that supports the deYelopment of the site specific ICs and EALs. This may include clarifications, references , e>rnmples , instructions fer calculations , etc. DeYeloper notes should not be included in the site's emergency classification scheme basis document.

Developers may elect to include information resulting from a deYeloper note action in a basis section. ECL l\ssignment Attributes Located within the Developer Notes section , specifies the attribute used fer assigning the IC to a given EGL. B3.4 IC AND EAL MODE APPLICABILITY The NEI 99 01 DAEC emergency classification scheme was developed recognizing that the applicability of I Cs and EALs will vary with plant mode. For example , some symptom-based ICs and EALs can be assessed only during the power operations, startup , or hot standby/shutdown modes of operation when all fission product barriers are in place , and plant instrumentation and safety systemsSAFETY SYSTEMS are fully operational.

In the cold shutdown and refueling modes, different symptom-based ICs and EALs will come into play to reflect the opening of systems for routine maintenance , the unavailability of some safety systemSAFETY 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 Recognition Category Mode AR C E F H s Power Operations X X X X X Startup X X X X X Hot Standby ' ' ' ' ' . . . . . Hot Shutdown X X X X X Cold Shutdown X X X X Refueling X X X X Defueled X X X X Permanently

' Defueled 20 NEI 99 0 I (Re~*isioA e) 1-Jo*,remaer 20 I 2 Typieal BWR DAEC Operating Modes Power Operations (1): Startup (2): Hot Shutdown (3): Cold Shutdown (4): Refueling (5): Mode Switch in Run Mode Switch in Startup/Hot Standby or Refuel (with all vesse l head closure bolts fu ll y tensioned)

Mode Switch in Shutdown , Average Reactor Coolant Temperature

>~212 °P (with all vessel head closure bolts fully tensioned)

Mode Switch in Shutdown , Average Reactor Coolant Temperatures

~212 °P (with all vessel head closure bolts fully tensioned)

Mode Switch in Shutdown or Refuel.,-ftflti (with one or more vessel head closure bolts less than fully tensioned}. Tvpieal P\\'R OPERi<\TING MODES Povyer Operations (I): Reactor Pov,cer > 5%, Keff > 0.99 Startup (2): Reactor Power~ 5%, Keff 0.99 Hot Standby (3): RCS~ 350 °f, Keff < 0.99 Hot Shutdovrn (4): 200 °f <RCS< 350 °f, Keff < 0.99 Cold Shutdown (5): RCS < 200 °f , Keff < 0.99 Refueling (6): One or more vessel head closure bolts less than fully tensioned Developers will need to incorporate the mode criteria from unit specific Technical Specifications into their emergency classification scheme. In addition , the scheme must also include the following mode designation specific to 1'lEI 99 01: Defueled (None): All fuel removed from the reactor vessel (i.e., full core offload during refueling or e J (tended outage). 21 NEI 99 QI (RevisioA

6) No>,*ember 2Q I 2 4 SITE SPECIFIC SCHEME DE\'EbOPMENT GUIDANCEDEVELOPMENT OF THE DAEC EMERGENCY CLASSIFICATION SCHEME This seetion provides detailed guidanee for developing a site speeifie emergeney elassifieation seheme. Coneeptually, the approaeh diseussed here mirrors the approaeh used to prepare emergeney operating proeedures generie material prepared by reaetor vendor ov,rners groups is eonverted by eaeh nuelear power plant into site speeifie emergeney operating proeedures. Likev,ise, the emergeney elassifieation seheme developer will use the generie guidanee in 1-lEI 99 01 to prepare a site speeifie emergeney elassifieation seheme and the assoeiated basis doeument.

It is important that the NEI 99 01 emergeney elassifieation seheme be implemented as an integrated paekage. Seleeted use of portions of this guidanee is strongly diseouraged as it ,viii lead to an ineonsistent or ineomplete emergeney elassifieation seheme that will likely not reeeive the neeessary regulatory approval.

4.1 GENERAL IMPLEMENTATION CUIDANCEDEVELOPMENT PROCESS The guidanee in NEI 99 01 is not intended to be applied to plants " as is"; however , developers should attempt to keep their site speeifie sehemes as elose to the generie guidanee as possible.

The goal is to meet the intent of the generie Initiating Conditions (ICs) and Emergeney Aetion Levels (EALs) within the eonteJ(t of site speeifie eharaeteristies loeale, plant design , operating features , terminology , ete. Meeting this goal will result in a shorter and less eumbersome NRG review and approval proeess , eloser alignment with the sehemes of other nuelear power plant sites and better positioning to adopt future industry wide seheme enhaneements.

When properly developed , the The DAEC ICs and EALs should were developed to be unambiguous and readily assessable.

As diseussed in Seetion 3 , the generie guidanee ineludes ICs and eJrnmple EALs. It is the intent of this guidanee that Q.Q!h be ineluded in site speeifie doeuments as eaeh serves a speeifie purpose. The IC is the fundamenta l event or condition requiring a declaration.

_The EAL(s) is the pre-determined threshold that defines when the IC is met. If some feature of the plant loeation or design is not eompatible with a generie IC or EAL , efforts should be made to identify an alternate IC or EAL. If an IC or EAL ineludes an explieit referenee to a mode dependent teehnieal speeifieation limit that is not applieable to the plant, then that IC and/or EAL need not be ineluded in the site speeifie seheme. In these eases , developers must provide adequate doeumentation to justify why the JC and/or EAL were not ineorporated (i.e., suffieient detail to allow a third party to understand the deeision not to ineorporate the generie guidanee).

Useful acronyms and abbreviations associated wit h the NEI 99 01DAEC emergency classification scheme are presented in Appendix A, Acronyms and Abbreviations. speeifie entries may be added if neeessary.

22 NEI 99 0 l (Re*,isioA

6) ~fovemeer

?Q l 2 Many words or terms used in the tffil 99 01DAEC emergency classification scheme have scheme-specific definitions.

These words and terms are identified by being set in all capital letters (i.e., ALL CAPS). The definitions are presented in Appendix B, Definitions.

Belov.* are e>rnmples of aeeeptable modifieations to the generie guidanee.

These ma)' be ineorporated depending upon site developer and user preferenees. The ICs within a Reeognition Category may be plaeed in reverse order for presentation purposes (e.g., start v,rith a General Emergeney at the left/top of a user aid , follov,*ed by Site Area E mergeney , Alert and NOUE). The Initiating Condition numbering may be ehanged. The first letter of a Recognition Category designation may be ehanged , as follows , provided the ehange is earried through for all of the assoeiated IC identifiers.

  • R may be used in lieu of A
  • M may be used in lieu of 8 For e>,ample, the Abnormal Radiation Levels/ Radiologieal Effluent eategory designator

" A" (for Abnormal) may be ehanged to " R" (for Radiation).

This means that the assoeiated ICs would be ehanged to RU 1, R1J2 , RA 1 , ete. The ICs and EALs from Reeognition Categories 8 and C may be ineorporated into a eommon presentation method (e.g., one table) provided that all related notes and mode applieability requirements are maintained. The ICs and EALs for Emergeney Direetorjudgment and seeurity related events may be plaeed under separate Reeognition Categories. The terms EAL and threshold may be used interehangeably. The material in the Developer Notes seetion is ineluded to assist developers with erafting eorreet IC and E AL statements.

This material is not required to be in the final emergeney elassifieation seheme basis doeument.

4.2 CRITICAL CHARACTERISTICS As diseussed above , developers are eneouraged to keep their site speeifie sehemes as elose to the generie guidanee as possible.

When crafting the scheme , developers should satisfy themselvesDAEC ensured that certain critical characteristics have been met. These critical characteristics are listed below.

  • The ICs, EALs, Operating Mode Applicabi l ity criteria, Notes and Basis information are consistent with industry guidance; while the actual wording may be different , the classification intent is maintained.

With respect to Recognition Category F, speeific scheme mustDAEC include~ some type of user-aid to facilitate timely and accurate classification of fission product barrier losses and/or potential losses. The user-aid logic must beis consistent with the classification logic presented in Section 9. 23 J>ffil 99 01 (ReYisioA a) J>Jo,*eFAber 2012

  • The ICs, EALs , Operating Mode Applicability criteria , Notes and Basis information are technically complete and accurate (i.e., they contain the information necessary to make a correct c l assification).
  • EAL statements use objective cr it eria and observable values.
  • ICs , EALs , Operating Mode App li cability and Note statements and formatting cons id er human factors and are user-friendly.
  • The scheme facilitates upgrading and downgrading of the emergency classification where necessary.
  • The scheme facilitates classification of multiple concurrent events or conditions. 24 l>ffil 99 01 (Re,*isioA 6) l>/oYefl'leer 2012 4.3 INSTRUMENTATION USED FOR EALs 4.4 Instrumentation referenced in EAL statements should include that described in the emergency plan section which addresses IO CFR 50.4 7(b)(8) and (9) and/or Chapter 7 of the FSAR. Instrumentation used for EALs need not be safety related , addressed by a Technical Specification or ODCM/RETS control requirement, nor powered from an emergency power source; ho,*, ever , EAL de>,'elopers should strive to DAEC incorporate g instrumentation that is reliable and routinely maintained in accordance with site programs and procedures.

Alarms referenced in EAL statements should beare those that are the most operationally significant for the described event or condition. Scheme developers should ensure that specified values used as EAL setpo int s are wit hin the calibrated range of the referenced instrumentation , and consider any automatic instrumentation functions that may impact accurate EAL assessment.

In addition, EAL setpoint va lu es should do not use terms such as " off-scale low" or " off-sca l e high" since that type of reading may not be readily differentiated from an instrument failure. Findings and violations related to EAL instrumentation issues may be located on the NRG v,rebsite.

PRESENTATION OF SCHEME INFORMATION TO USERS The US Nuclear Regulatory Commission (NRG) e>,pects licensees to establish and maintain the capability to assess, classify and declare an emergency condition promptly within 15 minutes after the availability of indications to plant operators that an emergency action level has been , or may be , e>rneeded. When writing an emergency classification procedure and creating related user aids , the developer must determine the presentation method(s) that best supports the end users by facilitating accurate and timely emergency cla s sification. To this end , developers should consider the following points. The first users of an emergency classification procedure are the operators in the Control Room. During the allowabl e classification time period , they may have responsibility to perform other critical tasks , and will likely have minimal assistance in making a classification assessment.

As an emergency situation evolves , members of the Control Room staff are likely to be the first personnel to notice a change in plant conditions. They can assess the changed conditions and , when warranted , recommend a different emergenC)'

classification level to the Technical Support Center (TSC) and/or Emergency Operations Facility (EOF). Emergency Directors in the TSC and/or EOF will have more opportunity to focus on making an emergency classification , and will probably have advisors from Operations available to help them. Emergency classification scheme information for end users should be presented in a manner with v,chich licensed operators are most comfortable.

Developers will need to work closely with representatives from the Operations and Operations Training Departments to develop readily usable and easily understood classification tools (e.g., a procedure and related user aids). If necessary, an alternate method for presenting 25 4.5 NEI 99 0 I (RevisioA G) }foyemeer 2012 emergeRcy classificatioR scheme iRformatioR may be developed for use by EmergeRcy Directors aRd/or Offsite RespoRse OrgaRizatioR persoRRel.

A wallboard is aR acceptable preseRtatioR method provided that it coRtaiRs all the iRformatioR Recessary to make a correct emergeRcy classificatioR.

This iRformation iRcludes the ICs, OperatiRg Mode Applicability criteria, EALs aRd Notes. Notes may be kept with each applicable EAL or moved to a commoR area aRd refereRced; a refereRce to a }fote is acceptable as loRg as the iRformatioR is adequately captured OR the wallboard aRd poiRted to by each applicable EAL+. Basis iRformatioR Reed Rot be iRcluded OR a wallboard but it should be readily available to emergeRcy classificatioR decisioR makers. IA some cases, it may be advaRtageous to develop two 1 Nallboards oRe for use duriRg power operatioRs, startup aRd hot coRditioRs, aRd aRother for cold shutdovm aRd refueliRg CORditiORS.

Alternative preseRtatioR methods for the RecognitioR Category F ICs aRd fissioR product barrier thresholds are acceptable aRd iRclude flow charts, block diagrams, aRd checklist type tables. Developers must eRsure that the site specific method addresses all possible threshold combiRatioRs aRd classification outcomes shovm iR the BWR or PWR EAL fissioR product barrier tables. The NRG staff coRsiders the preseRtatioR method of the RecogRitioR Category F iRformation to be aR importaRt user aid aRd may request a chaRge to a particular proposed method if, amoRg other reasoRs, the chaRge is Recessary to promote coRsisteRcy across the iRdustry.

INTEGRATION OF ICs/E,<\l,s WITII PLANT PROCEDURES i\ rigorous iRtegratioR of IC aRd EAL refereRces iRto plaRt operatiRg procedures is Rot recommeRded.

This approach 'Nould greatly iRcrease the admiRistrative coRtrols aRd workload for maiRtaiRiRg procedures.

OR the other haRd, performaRce challeRges may occur if recogRitioR of meetiRg aR IC or EAL is based solely on the memory of a liceRsed operator or aR EmergeRcy Director, especially duriRg periods of high stress. Developers should coRsider placiRg appropriate visual cues (e.g., a step, note, cautioR, etc.) iR plaRt procedures alertiRg the reader/user to coRsult the site emergeRcy classificatioR procedure.

Visual cues could be placed iR emergeRcy operatiRg procedures, abRormal operatiRg procedures, alarm respoRse procedures, aRd Rormal operatiRg procedures that apply to cold shutdowR aRd refueliRg modes. As aR e~(ample, a step, Rote or cautioR could be placed at the begiRRiRg of aR RCS leak abRormal operatiRg procedure that remiRds the reader that an emergeRcy classificatioR assessmeRt should be performed. Where appropriate , the Notes shovm iA the geAerie g1:1iaaAee typieally iAel1:1Ele the eveAtleoAaitioA EGL aml the E11:1ratioA time speeifiea iA the EAL. If Elevelopers prefer to have seyeral IGs refereAee a eommoA l>l"OTE OR a v,*alleoara aisplay, it is aeeeptaele to remo*,e the EGL aAEI time eriterioR aAa 1:1se a geAerie statemeAt.

For ellample, a eommoA NOTE eo1:1IEI reaa " The EmergeAey Direetor sho1:1la aeelare the emergeAe)'

promptly 1:1poA EletermiAiAg that the applieasle EAL time has '3eeA e1(eeeaea , or will lil,ely '3e eMeeeaea." 26 4.6 BASIS DOCUMENT ~J EI 99 0 I (ReyisioR

6) NoYember 2012 A basis document i s an integral part of an emergency class i fication scheme. The mater i al in this document supports proper emergency c l assification decision making by providing informing background and deve l opment information in a readi l y accessib l e format. It can be referred to in training situations and when making an actual emergency classification , if necessary.

The document is also useful for establishing configuration management controls for EP related equipment and e1(plaining an emergency classification to offsite authorities. The content of the basis document should include , at a minimum , the following:

  • A site specific Mode Applicab ili ty Matri1( and descript i on of operating modes , similar to that presented in section 3.5.
  • A discussion of the emergency classification and declaration process reflecting the material presented in Section 5. This materia l may be edited as needed to align v,*ith site specific emergency plan and implementing procedure requirements.
  • Each Initiating Condition along 1,vith the associated EALs or fission product barrier thresholds , Operating Mode Applicability , *Notes and Basis information.
  • A listing of acronyms and defined terms , simi l ar to that presented in Appendices A and B , respectiYely. This material may be edited as needed to align with site specific characteristics.
  • Any site specific background or technical appendices that the deve l opers believe would be useful in e1(plaining or using elements of the emergency classification scheme. 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 readi l y available to be referenced , if necessary , by the Emergency Director. For e1rnmple , a copy of the basis document could be ma i ntained in the appropriate emergency response facilities. Because the information in a basis document can affect emergency classification decision making (e.g., the Emergency Director refers to it during an event), the NRG staff expects that changes to the basis document v,ill be evaluated in accordance with the provisions of 10 GFR 50.54(q). 4,+4.4 EAL/THRESHOLD REFERENCES TO AOP AND EOP SETPOINTS/CRITERIA As reflected in the generic guidance,Some of the criteria/values used in several EALs and fission product barrier thresholds may be ar e drawn from a plant'sDAEC AOPs and EOPs. This approach is intended to maintain good alignment between operational diagnoses and emergency classification assessments. Developers should verify that aA ppropriate administrative controls are in place to ensure that a sub s equent change to an AOP or E OP i s screened to determine if an evaluation pursuant to 10 CFR 50.54(q) is required.

27 4.8 DEVELOPER AND USER FEEDBACI( --------------------~ ~ffil 99 Q 1 (Re,,*isioR 6) NoYemeer 2012 Questions or eomments eoneeming the material in this doeument may be direeted to the }ffil Emergeney Preparedness staff , }ffil EAL task foree members or submitted to the Emergeney Preparedness Frequently Asked Questions proeess. 2 8 NE I 99 01 (R e\'i s ioA 6) N o Ye mb e r 201 2 5 GUIDANCE ON MAKING EMERGENCY CLA&&IFICATION&USING THE DAEC EALS 5.1 GENERAL CONSIDERATIONS When making an emergency class ifi cation , the Emergency Director must co n s id er a ll information having a bearing on the proper assessment of an Initiating Cond iti on (IC). This includes the Emergency Action Level (EAL) plus the assoc i ated Operating Mode Applicabi lit y , Notes a nd the informing Basis information.

In the Recognition Category F matrices , EALs are referred to as Fission Product Barrier Threshold s; the thresholds serve the same function as an EAL. NRC regulations require the licensee to estab li s h and maintain the capability to assess , classify, a nd declare an emergency condition within 15 minutes after the ava ilabili ty of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency cond iti on as soon as possible fo ll owing identification of the appropriate emerge nc y classification l eve l. The NRC staff ha s provided g uid a n ce on implementing this requirement in NSIR/DPR-ISG-01 , Interim Staff Guidance, Em e rg e ncy Planning for Nuclear Power Plant s. All emergency classification assessments shou ld be based upon valid indications , reports or cond ition s. A valid indic atio n , report , or condition , is o n e that ha s been verified through appropriate means such that there is no doubt regarding the indicator's operabi lit y , the condit i on's existence , or the report's accuracy. For examp l e , validation could be accomplished through an instrument c hann e l check , response on r e l ated or redundant indicators , or direct observat ion by plant personnel.

The va lid at i on of indic atio ns should be comp l eted in a manner that s upp orts timely emerge ncy declaration. For ICs and EALs that have a s tipu l ated time duration (e.g., 15 minutes , 30 minutes , etc.), the Emergency Director sho uld not wait until the applicable time has e lap sed, but sho uld declare the event as soo n as it is determined that the conditio n has exceeded , or will likely exceed , the app li cable time. If an ongo in g radiological release i s detected and the release start time is unknown , it should be assumed that the release duration specified in the IC/E AL has been exceeded , absent data to the contrary. A planned work activ it y that resu lt s in an expected event or con diti on wh i ch meets or exceeds an EAL does not warrant an emergency declaration provided that l) the act i v i ty proceeds as planned and 2) the p l ant remains w ithin the limits imp osed by the operating li cense. S uch activities include planned work to test , manipulate , repair, maintain or modify a system or component.

In these cases , the controls associated w ith the planning , preparation and exec ution of the work will ens ure that comp lianc e i s maintained with all aspects of the operating li cense provided that the act i vity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 10-§-CFR 50.72. The assessment of some EALs i s ba sed on the results of analyses that are nec essary to ascertain whether a specific EAL threshold has been exceeded (e.g., dose assessments, 29

}'JEI 99 QI (Re,*isioA <:i) }'lo,*ember 2Q 12 chemistry sampling, RCS leak rate calculation, 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 expects licensee s to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).

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 ba se d on operator/management experience and judgment is still necessary.

The NEI 99 -0+ This scheme provides the E mergency Dir ecto r 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 E mergency Director 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. 5.2 30 CLASSIFICATION METHODOLOGY NEl 99 0 I (Re\*isioA

6) No 1 ,*ember 2012 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/DPRISG-01. 5.3 CLASSIFICATION OF MULTIPLE EVENTS 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 one unit or at t>.vo different units , a Site Area Emergency should be declared. +Additionally.

t here is no " additive" effect from multiple EALs meeting the same ECL. For example: If two Alert EALs are met., , whether at one unit or at two different units , 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 ofNRC Guidance for Emergency Notifications During Quickly Changing Events. 31 5.4 CONSIDERATION OF MODE CHANGES DURING CLASSIFICATION NEI 99 Q 1 (Re;cisioR

6) ~lo;cember 2012 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 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 event 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.

For events that occur in Cold Shutdown or Refueling , escalation is via EALs that are applicable in the Cold Shutdown or Refueling modes , even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response.

In particular , the fission product barrier E ALs are applicable only to events that initiate in the Hot Shutdown mode or higher. 32 5.5 CLASSIFICATION OF IMMINENT CONDITIONS NEI 99 Ql (Re11isioA

6) l>lo*,*ember 2Q 12 Although EALs provide specific thresholds , the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT).

If , in the judgment of the Emergency Director , meeting an EAL is IMMINENT , the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels , this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.

5.6 EMERGENCY CLASSIFICATION LEVEL UPGRADING AND DOWNGRADING An ECL may be downgraded when the event or condition that meets the highest IC and EAL no longer exists, and other site-specific downgrading requirements are met. If downgrading the ECL is deemed appropriate , the new ECL would then be based on a lower applicable IC(s) and EAL(s). The ECL may also simply be terminated.

33 I s.1 NEI 99 01 (Re't'isioA

6) }fo,,cemaer 2012 The following approach to downgrading or terminating an ECL is recommended.

ECL Action When Condition No Longer Exists Unusual Event Terminate the emergency in accordance with plant procedures.

Alert Downgrade or terminate the emergency in accordance with plant procedures.

Site Area Emergency with no Downgrade or terminate the emergency in long-term plant damage accordance with plant procedures.

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.

As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02. 34 CLASSIFICATION OF SHORT-LIVED EVENTS NEJ 99 0 l (RevisioR

6) No*,cember 20 I? As discussed in Section 3.2, event-based I Cs 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.

If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration.

Examples of such events include a failure of the reactor protection system to automatically scram~ the reactor followed by a successful manual scram~ or an earthquake.

5.8 CLASSIFICATION OF TRANSIENT CONDITIONS Many of the I Cs 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 few seconds to a few minutes).

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.

35 J>JEI 99 QI (ReYisioR

6) November 2012 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 successfu l in correcting the condit ion 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 fai l s to automatically start. Steam generator levels rapid l y 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 c l ears the inadequate RCS heat removal condition prior to an emergency declaration , then the classificat ion shou ld be based on the A TWS on l y. It is important to stress that the 15-minute emergency classification assessment period i s 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 delays. The provision discussed above addresses only those rapidly evolving s itu ations where an operator is able to take a successful corrective action prior to the Emergency Director 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 cond iti ons. 5.9 AFTER-THE-F ACT 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-1022 is applicable.

Specifically, the event shou ld be reported to the NRC in accordance with 10 CFR-§-50.72 w ithin 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.

5.10 RETRACTION OF AN EMERGENCY DECLARATION Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022.

36 Ne! 99 Q 1 (RevisioR

6) Deeemaer 2Q IQ 6 ABNORMAL RAD LEVELS/ RADIOLOGICAL EFFLUENT ICS/EALS Table AR 1: Reeegnitien Categerv "AR" Initiating Cenditien Matrix UNUSUAL EVENT AUlRUl Release of gaseous or liquid radioactivity greater than 2 times the (site specific effluent release controlling document)ODCM limits for 60 minutes or longer. Op. },fedes: All UNPLA1'il'ffiD loss of water level above irradiated fuel. Op. },fedes: All ALERT 1 A....A1R,A,J Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid CDE. Op. },fedes: All AA2RA2 Significant lowering of water level above, or damage to , irradiated fuel. Op. ,\fedes: All SITE AREA. EMERGENCY AS1RS1 Release of gaseous radioactiYity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid CDE. Op. },fedes: All Spent fuel pool level at (site specific Level 3 description)1Q ft 8 in (Level 3}. ,AnA,.J!MJ

-Radiation levels that impede access to equipment necessary for normal plant operations , cooldown or shutdown.

Op. Afedes: All 3 7 GENERAL EMERGENCY AGlRGl Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid CDE. Op. },fedes: AG2RG2 Spent fuel pool level cannot be restored to at least (site specific Level 3 description)4 0 ft 8 in (Level 3) for 60 minutes or longer. All -

Nel 99 g l (ReYisioA

6) NO\'efl'leer 2Q 12 AU1RU1 ECL: Notification of Unusual Event Initiating Condition:

Release of gaseous or liquid radioactivity greater than 2 times the specific effluent re l ease contro ll ing document)ODAM limits for 60 minutes or longer. Operating Mode Applicability:

All Emergency Action Levels: Example Emergeeey ,A_._etiee Le11els: (1 or 2 or 3) Notes: *

  • The Emergency Director should declare the Unusual Eventevent promptly upon determining that the applicab l e time60 m i nutes has been exceeded , or will l ikely be exceeded.

If an ongoing release is detected and the release start time is unknown , assume that the re l ease duration has exceeded the specified time limit60 minutes. If the effluent flow past an effluent mon i tor is known to have stopped due to act i ons to iso l ate the release path , then the effluent monitor reading is no longer valid for classification purposes.

Reading on ANY Tab l e R-1 effluent radiation monitor greater than colu mn "NOUE" for+/- times the (site specific effluent release controlling document) limits for 60 minutes or lon ger: Monitor NOUE Reactor Building ventilation rad 8.0E-04 uci/cc monitor (Kaman 3/4, 5/6, 7/8} V> Turbine Building ventilation rad 8.0E-04 uci/cc ::J monitor (Kaman 1/2} 0 <lJ V> Off gas Stack r ad monitor ro 2.0E-01 uci/cc t!) (Kaman 9/10} L LRPSF rad mon i tor 1.2E-03 uci/cc (Kaman 12} GSW rad monitor 1.SE+03 cp s (RIS-4767} ) RHRSW & ESW rad monitor 8.4E+02 cp s (RM-1997} RHRSW & ESW Rupture Di s c rad 1.0E+03 cps monitor (RM-4268}

R 1.2+/- (site specific monitor list and threshold values corresponding to 2 times the controlling document limits) R ading on AN¥ANY effluent radiation monitor greater than 2 times the a l arm setpoint established by a current radioactivity discharge permit for 60 minutes or longer. 38

~3 },ffil 99 QI (Re\*isioH 6) },lo\*emeer 2012 Sample analysis for a gaseous or liquid release indicates a concentration or release rate greater than 2 times the (site specific effluent release controlling document)ODAM limits for 60 minutes or longer. 39 Definitions:

Basis: NBI 99 QI (Re\*isioR e) November 2Q 12 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 (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release , monitored or monitored , including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plantsDAEC incorporates 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 E ALs 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. Classification based on effluent monitor readings assumes that a release path to the environment is established.

_I f 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 RUl.1 -This EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways. EAL RUl.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 will typically be associated with planned batch releases from non-continuous release pathways (e.g., rad waste , waste gas). EAL RUl .3 -This EAL addresses uncontrolled gaseous or liquid releases that are detected by sample analys el s or environmental surveys , particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains , heat exchanger leakage in river water systems , etc.). Escalation of the emergency classification level would be via IC AA-l-RA 1. Developer Notes: The "site specific effluent release controlling document" is the Radiological Effluent Technical Specifications (RET8) or, for plants that have implemented Generic Letter 89 Ol-1-,--tfle

+ fmp,lemenlr:1titm

£>/ Pregrammatie Centre ls fer Radielegieal Effluent Teehniea!

Speeijieatiens in the Administrati*;e Centrels Seetien &/the Teehniea!

Speeijieatiens a19d the Releeatien

&j Preeedura!

Details &JRETS le the Of!site Dese Ceteulatien J.lam1al er te the Preeess Centre! Pregram 40 01 (Re*,*isioA

6) l>ffil 99 l mber 2012 } OYe

~!El 99 QI (RevisioR

6) Noyemeer 2Q 12 scope of the plant Technical Specifications.

A licensee may request to include an EA~ using real time dose projection system results; approval ,,viii be considered on a case by case basis. Indications from a perimeter monitoring system are not included ~n the generic EA.Ls. Many licensees do not have this capability.

For those that.do.' these monitors may not be co~trolled and maintained to the same level as plant equipment , or ,v1thm ~he scope of the plant Techm:al Specifications.

In addition , readings may be influe~ced by enY_1ro~mental or ~ther f~~to 1~~: ,

  • licensee may request to include an EAL using a penmeter mon1tormg system , approval "Ill be considered on a case by case basis. E GL Assignment Attributes:

3 .1.1.B 42

~IEI 99 0 I (RevisioR

6) November 2012 AU2RU2 ECL: Notification of Unusual Event Initiating Condition:

UNPLANNED loss of water level above irradiated fuel. Operating Mode Applicability:

All l E*emple Emergency Action Levels: + a. UNPLANNED water level drop in the REFUELING PATHWAY as indicated by ANY of the following:

  • Report to control room (visual observation)
  • Fuel pool level indication (LI-3413)

LESS THANless than 36 feet and lowering

  • WR GEMAC Floodup indication (LI-4541) coming on scale(site specific level indications).

AND b. UNPLANNED rise in area radiation levels as indicated by ANY of the following radiation monitors.

  • (site specific list of area radiation monitors)

Spent Fuel Pool Area, Rl-9178

  • North Refuel Floor, RI-9163
  • New Fuel Vault Area, RI-9153
  • South Refuel Floor, Rl-9164
  • NW Drywell Area Hi Range Rad Monitor, RIM-9184A
  • South Drywell Area Hi Range Rad Monitor. RIM-9184B Definitions:

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. REFUELING PATHWAY: The reactor refueling cavity. spent fuel pool and fuel transfer canal. 43

~ Basis: NEI 99 QI (Re*,isioA

6) ~loyemaer 2Q 12 This IC addresses a decrease in water l evel above i rradiated fuel sufficient to cause elevated radiation levels. This condit i on co uld be a precursor to a more serious event and is also indicative of a minor loss in the ability to control rad i at i on levels within the plant. It is therefore a potential degradation in the l evel of safety of the plant. A water level decrease will be primarily determined by indications from available l evel instrumentation.

_Other sources of level indications may in clude repo1ts from plant personnel (e.g., from a refueling 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 mon it ors in those l ocations.

44

~JEI 99 01 (Re,,*isioA e) ~fovemeer 2012 The effects of planned evolutions should be considered.

For examp l e, a refueling bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vesse l head or movement of a fuel assemb l y. Note that this EAL is app l icable only in cases where the elevated reading is due to an UNPLANNED loss of water level. During preparation for reactor cavity flood up prior to entry into refuel mode, reactor vessel level instrument LI-4541 (WR GEMAC, FLOODUP) on control room panel 1C04 is placed in service by I&C personnel connecting a compensating air signal after the reference leg is disconnected from the reactor head. Normal refuel water level is above the top of the span of this flood up level indicator. A valid indication (e.g., not due to loss of compensating air signal or other instrument channel failure) ofreactor cavity level coming on span for this instrument is used at DAEC as an indicator of uncontrolled reactor cavity level decrease.

DAEC Technical Specifications require a minimum of 36 feet of water in the spent fuel pool when moving irradiated fuel into the secondary containment.

During refueling, the gates between the reactor cavity and the refueling cavity are removed and the spent fuel pool level indicator LI-3413 is used to monitor refueling water level. Procedures require that a normal refueling water level be maintained at 37 feet 5 inches. A low level alarm actuates when spent fuel pool level drops below 37 feet 1 inch. Symptoms of inventory loss at DAEC include visual observation of decreasing water levels in reactor cavity or spent fuel storage pool, Reactor Building (RB) fuel storage pool radiation monitor or refueling area radiation monitor alarms, observation of a decreasing trend on the spent fuel pool water level indicator, and actuation of the spent fuel pool low water level alarm. To eliminate minor level perturbations from concern, DAEC uses LI-3413 indicated water level below 36 feet and lowering.

Increased radiation levels can be detected by the local area radiation monitors surrounding the spent fuel pool and refueling cavity areas. Applicable area radiation monitors are those listed in AOP 981. A drop in water l evel above irradiated fuel within the reactor vesse l may be class ifi ed in accordance Recognition Category C during the Cold Shutdown and Refueling modes. Escalation of the emergency classification level wou ld be via IC ~RA2. DeYelapeF Notes: The " site specific le1t*el iAdicatioAs" are those iAdicatioAs that may be used to moAitor water )eye) iA the Yarious portioAs of the REFUELING PATHWAY. Specify the mode applicability of a particular iAdicatioA if it is Aot a*,ailable iA all modes. The "site specific list of area radiatioA moAitors" should coAtaiA those area radiatioA moAitors that would be e~,pected to have iAcreased readiAgs follov,iAg a decrease iA water level iA the site specific REFUELR>JG PATHWAY. lA cases where a radiatioA moAitor(s) is AOt a1t*ailable or would Aot provide a useful iAdicatioA , coAsideratioA should be giveA to iAcludiAg alternate iAdicatioAs such as UNPLAl>n>lED chaAges iA taAk aAd/or sump levels. DevelopmeAt of the EALs should coAsider the availability aAd limitations of mode depeAdeAt , or other coAtrolled but temporary , radiatioA moAitors. Specify the mode applicability of a particular moAitor if it is Aot available iA all modes. EGL AssigAmeAt Attributes:

3 .1.1.A aAd 3 .1.1.B 45 ECL: Alert NEI 99 0 I (Re*,'isioA

6) No¥e1l'!eer 2012 AA1RA1 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: ExamtJle Emergeney f ... etion Le1,*els:

I

  • The Emergency Director should declare the Alert event promptly upon determining that the applicable time has been exceeded , or will likely be exceeded.
  • I
  • T l I I I I I I I I I If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limitl 5 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 l J should Q!!ly_be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

Reading on ANY of the followingTable R-1 effluent radiation monitor s greater than the reading shovmcolumn "Alert" for 15 minutes or longer: V, ::::, 0 (]J V, ro t!) Moni t or Reactor Building ventilation rad monitor {Kaman 3/4, 5/6, 7 /8} Turbine Building ventilation rad monitor {Kaman 1/21 Off gas Stack rad monitor {Kaman 9/101 LLRPSF rad monitor {Kaman 121 GSW rad monitor {RIS-47671 RHRSW & ESW rad monitor {RM-1997}

RHRSW & ESW Rupture Disc rad monitor {RM-4268} 1.4E-02 uci/cc 4.SE+Ol uci/cc 1.4E-02 uci/cc 1.7E+04 cps 1.2E+04 cps 1.8E+04 cps 46 (site specific monitor list and threshold values) NEI 99 0 I (RevisioA e) ~fo .. *emaer 2012 se assessment using actual meteorology indicates doses greater than 10 mrem TEDE ____ or 50 mrem thyroid CDE at or beyond (site specific dose receptor point)SITE UNDARY. [Preferred]

Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses greater than 10 mrem TEDE or 50 rnrem thyroid CDE at or beyond fstte-specific dose receptor point)the SITE BOUNDARY for one hour of exposure. R A 1.4 Field survey results indicate EITHER of the following at or beyond (site specific dose receptor point)the SITE BOUNDARY:

  • Closed window dose rates greater than 10 mR/hr 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. 47 Definitions:

NEI 99 QI (RevisioR e) No, 1 effieer 2012 SITE BOUNDARY:

That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

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 (PA Gs). 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 significant uncontrolled release).

This JC is modified by a note that EAL RA 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions.

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 COE 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.

_I f 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 AS+RS 1. Develeper Netes: While this IC may not be met absent challenges to one or more fission product barriers , it provides classification diversity and may be used to classify events that would not reach the same EGL based on plant status or the fission product matri>, alone. For many of the DB As analyzed in the Updated Final Safety Analysis Report , the discriminator will not be the number of fission product barriers challenged, but rather the amount of radioactivity released to the environment.

The EPA PAGs are e>cpressed in terms of the sum of the effective dose equivalent (EDE) and the committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent (CDE). For the purpose of these IC/EALs , the dose quantity total effective dose equivalent (TEDE), as defined in 10 CFR § 20 , is used in lieu of" ... sum of EDE and CEDE .... ". The EPA PAG guidance provides for the u s e of adult thyroid dose conversion factors; ho\vever, some states have decided to base protective actions on child thyroid CDE. Nuclear power 48 NEI 99 Ql (ReYisieA e) Ne*ref1'1'3er 2Ql 2 plant ICs/EALs need to be consistent with the protective action methodologies employed by the States within their EPZs. The thyroid CDE dose used in the JC and EALs should be adjusted as necessary to align with State protective action decision making criteria.

The " site specific monitor list and threshold values" should be determined with consideration of the following:

  • Selection of the appropriate installed gaseous and liquid effluent monitors.
  • The effluent monitor readings should correspond to a dose of IO mrem TEDE or 50 mrem thyroid CDE at the " site specific dose receptor point" (consistent with the calculation methodology employed) for one hour of e>1.posure.
  • Monitor readings v., ill be calculated using a set of assumed meteorological data or atmospheric dispersion factors; the data or factors selected for use should be the same as those employed to calculate the monitor readings for ICs ASI and AGI. Acceptable sources of this information include , but are not limited to , the RETS/ODCM and values used in the site's emergency dose assessment methodology.
  • The calculation of monitor readings v,ill also require use of an assumed release isotopic mix; the selected mi>1. should be the same as that employed to calculate monitor readings for ICs ASl and AGI. Acceptable sources of this information include , but are not limited to , the RETS/ODCM and values used in the site's emergency dose assessment methodology.
  • Depending upon the methodology used to calculate the EAL values , there may be overlap of some values between different ICs. Developers will need to address this overlap by adjusting these values in a manner that ensures a logical escalation in the EGL. The " site specific dose receptor point" is the distanee(s) and/or locations used by the licensee to distinguish between on site and offsite doses. The selected distanee(s) and/or locations should reflect the content of the emergency plan, and the procedural methodology used to determine offsite doses and Proteeti*,ce Action Recommendations.

The variation in selected dose receptor points means there may be some differences in the distance from the release point to the calculated dose point from site to site. Developers should research radiation monitor design documents or other information sources to ensure that 1) the EAL value being considered is within the usable response and display range of the instrument , and 2) there are no automatic features that may render the monitor reading invalid (e.g., an auto purge feature triggered at a particular indication level). It is recognized that the condition described by this IC may result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. In those eases , EAL values should be determined 1 with a margin sufficient to ensure that an accurate monitor reading is available.

For e>rnmple , an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identif')r an alternate EAL threshold. Although the IC references TEDE , field survey results are generally available only as a " whole body" dose rate. For this reason , the field survey EAL specifies a " closed 1 ,vindow" survey reading. Indications from a real time dose projection system are not included in the generic EALs. 49 1>ffil 99 QI (Re,,isioR e) No 1 remeer 2912 Many licensees do not haYe this capability.

For those that do , the capability may not be within the scope of the plant Technical Specifications.

A licensee may request to include an EA~ using real time dose projection system results; approYal will be considered on a case by case basis. Indications from a perimeter monitoring system are not included in the generic EALs. Many licensees do not have this capability.

For those that do , these monitors may not be co_ntrolled and maintained to the same leYel as plant equipment , or within the scope of the plant Technical Specifications.

In addition , readings may be influenced by environmental or other factors. A licensee may request to include an EAL using a perimeter monitoring system; approYal 'will be considered on a case by case basis. EGL Assignment Attributes:

3.1.2.C 50 ECL: Alert NEI 99 g l (Re\*isieA e) Ne;rember 2012 AA2RA2 Initiating Condition:

Significant lowering of water level above , or damage to , irradiated fuel. Operating Mode Applicability:

All Emergency Action Levels: E1rnmple Emergency Action Levels: (1 or 2 or 3) Uncovery of irradiated fuel in the REFUELING PATHWAY. Damage to irradiated fuel resulting in a release of radioactivity from the fuel as indicated by ANY of the follo 1.ving radiation monitors::Hi Rad alarm for ANY of the following ARMs:

  • Spent Fuel Pool Area, RI-9178
  • North Refuel Floor, RI-9163
  • New Fuel Vault Area, RI-9153
  • South Refuel Floor, RI-9164 Reading greater than 5 R/hr on AN¥ANY of the following radiation monitors (in Mode 5 only):
  • NW Drywell Area Hi Range Rad Monitor, RIM-9184A
  • South Drywell Area Hi Range Rad Monitor, RIM-9184B RA2.3 (site specific listing of radiation monitors , and the associated readings , setpoints and/or alarms) L wering of spent fuel pool level to (site specific Level 2 value). [Sec Dcvclepcr Netcs]25.

l 7 feet. Definitions:

REFUELING PATHWAY -The reactor refueling cavity, spent fuel pool and fuel transfer canal. Basis: 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 (sec Dc 1;clepcr ]\fetcs). 51 "t-ffil 99 0 I (RevisieA e) "t-fo*,ember 2012 These events present radio l ogical safety cha ll enges 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 l eve l of safety of the plant. Expected radiation monitor alarm(s) during preplanned transfer of highly radioactive material through the affected areas are not considered valid alarms for the purpose of comparison to these EALs. 52

}ffil 99 Q 1 (RevisioA

6) }fo*remeer 2Q 12 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-HUl. Esca lation of the emergency would be based on either Recognition Category A-R or CI Cs. EAL RA2.l This EAL escalates from ~RU2 in that the los s of level , in the affected portion of the REFUELING PATHWAY , is of sufficient magnitud e to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters.

Computational aids may also be used (e.g., a boil off curve). Classification of an event using this EAL should be based on the totality of available indications , reports , and observations.

53 J>ffil 99 0 I (RevisioA e) J>fo\*emeer 2012 While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUELING PATHWAY , the reading may not be a reliable indication of whether or not the fuel i s 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 RA2.2 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 readingsAn alarm on these radiation monitors should be considered in conjunction with in-plant reports or observations of a potenti a l fuel damaging event (e.g., a fuel handling accident).

Threshold values for the Drywell monitors are only applicable in Mode 5 since the calculated radiation levels from damage to irradiated fuel would be masked by the typical background levels on these monitors during plant operation, and mechanical damage to a fuel assembly in the vessel can only happen with the reactor head removed. EAL RA2.3 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. Esca lation of the emergency classification level would b e via ICs AS+-RS 1 or AS+/--RS2(see AS2 De11eleper

]Vetes). Developer Notes: For EAL #1 Depending upon the availability and range of instrumentation , this EAL may include specific readings indicative of fuel uncovery; consider water and radiation level readings.

Specify the mode applicability of a particular indication if it is not available in all modes. For EAL #2 The " site specific listing of radiation monitors , and the associated readings , setpoints and/or alarms" should contain those radiation monitors that could be used to identify damage to an irradiated fuel assembly (e.g., confirmatory of a rel e ase of fission product gases from irradiated For EALs #1 and #2 Developers should research radiation monitor design documents or other information sources to ensure that 1) the EAL value being considered is within the usable response and display 54 NEI 99 01 (RevisioR

6) November 2012 range of the instrument , and 2) there are no automatic features that may render the monitor reading invalid (e.g., an auto purge feature triggered at a particular indication level). It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases , EAL values should be determined 1 uith a margin sufficient to ensure that an accurate monitor reading is available.

For example , an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold. To further promote accurate classification , developers should consider if some combination of monitors could be specified in the EAL to build in an appropriate level of corroboration between monitor readings into the classification assessment.

Development of the EALs should also consider the availability and limitations of mode dependent , or other controlled but temporary , radiation monitors.

Specify the mode applicability of a particular monitor if it is not available in all modes. For EAL #3 In accordance with the discussion in Section 1.4 , NRG Order EA 12 051 , it is recommended that this EAL be implemented when the enhanced spent fuel pool level instrumentation is available for use. The " site specific Level 2 value" is usually the spent fuel pool level that is adequate to provide substantial radiation shielding for a person standing on the spent fuel pool operating deck. This site specific level is determined in accordance with NRG Order EA 12 051 and NEl 12 02 , and applicable owner's group guidance.

Developers should modify the EAL and/or Basis section to reflect any site specific constraints or limitations associated 1,vith the design or operation of instrumentation used to determine the Level 2 value. EGL Assignment Attributes:

3.1.2.B and 3.1.2.C 55 ECL: Alert NEl 99 01 (Re1,*isioH

6) No,*emaer 2012 AA3RA3 Initiating Condition:

Radiation levels that impede access to equipment areas necessary for normal plant operation s, cooldown or shutdov~*n. Operating Mode Applicability:

All ---Emergency Action Levels: Example Emergeeey Aetioe Levels: (1 or 2) Note: If the equipment in the listed room or area ,vas already inoperable or out of service before the e11ent occurred, then no emergency classification is warranted.

R 3.1 Dose rate greater than 15 mR/hr in ANY of the following areas:

  • Control Room ARM-(RM-9162)
  • Central Alarm Station (by survey) (other site specific areas/rooms) the following plant rooms or areas: (site specific li st of plant rooms or areas with entry related mode applicability identified)

Definitions:

Basis: 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 Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.

56 NEI 99 QI (Re\*isieR

6) J,,foyemaer 2Q 12 For EAL 2 , an Alert declaration is 1,varranted 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 e>ltraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding , requiring use of non routine protective equipment , requesting an e>ltension in dose limits beyond normal administrative limits). An emergency declaration is not vrarranted 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 e>rnmple , the plant is in Mode 1 1 n'hen the radiation increase occurs , and the procedures used for normal operation , cooldovm 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 1,vhich address the temporary inaccessibility of a room or area (e.g., radiography , spent filter or resin transfer , etc.). The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

The access control measures are of a conservative or precautionary nature, and 'n'ould not actually prevent or impede a required action. Esca l a ti o n of t h e e m erge n cy c l assificat i o n l eve l wo ul d b e v i a Recogn i tio n Catego r y AR , C or F ICs. De\*elopeF Notes: EAL#l The valu e of l 5mR/hr is derived from the GDC 19 value of 5 rem in 30 days with adjustment for e>lpected occupancy times. The " other site specific areas/rooms" should include any areas or rooms requiring continuous occupancy to maintain normal plant operation , or to perform a normal cooldovm and shutdown.

EAL#2 The "site specific list of plant rooms or areas with entry related mode applicability identified" should specify those room s or areas that contain equipment which require a manual/local action as specified in operating proc e dures used for normal plant operation , cooldown and shutdovm.

Do not include rooms or areas in which actions of a contingent or emergency nature would be performed. (e.g., an action to address an off normal or emergency condition such as emergency repairs , corrective measures or emergency operations). In addition , the list should specify the plant mode(s) during which entry '.Vould be required for each room or area. 57 NEI 99 0 I (RevisioA

6) Novemeer 2012 The list should not include rooms or areas for 1 n<hich entr)' is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

If the equipment in the listed room or area 'Nas already inoperable , or out of service , before the event occurred , then no emergency should be declared since the event will have no adverse impact beyond that already allowed by Technical Specifications at the time of the event. Rooms and areas listed in EAL #1 do not need to be included in EAL #2, including the Control Room. EGL Assignment Attributes:

3.1.2.C 58 J>JEl 99 0 I (Re*,ision

6) J>lovemeer 2012 A S1RS 1 ECL: S it e Area Emergency Initiating Condition:

Release of gaseo u s radioactivity resulting in off s it e dose greater than 100 mrem TED E or 500 mrem thyroid CDE. Operating Mode Applicability:

All Emergency Action Levels: ExamJlle Emergeney ,A.._etien LeYels: (1 or 2 or 3) Notes: I * *

  • I
  • T h e Emergency Director s h o uld declare the Site Area Emergeneyevent promptly up on determining that the app li cab l e time h as b een exceede d , or w ill lik e l y b e excee d ed. lf a n ongo in g release is detected and the release start time is unknown , assume that the release duration has exceeded 15 minutesthe spec i fied time limit. If the effluent flow past a n effluent monitor is known to have stopped due to actions to i so l ate the release path , then the effl uent monitor reading i s n o l onger va lid for c l ass ifi cat i on purposes.

The pre-calculated efflue nt monitor values presented in EAL L I s h o uld_QJy b e u sed for e m ergency c l assificatio n assessments until the results from a dose assessment u s in g act u a l meteorology are availab l e. Reading on ANY of the followingTab l e R-1 effluent radiat ion monit or s greater than co l umn "SAE" the reading shown for I 5 minut es or l o n ger: j V, ::J 0 QJ V, rtl l!) Effluent Meniter la

  • Reaetor Ih1ileing Yent il ation rae 11'1onitor (Kaman 3/4, 5/6, 7/8) T1:1reine B1:1ileing Yentilation rae ll'IOnitor (Kaman 1/2) Offuas Staek rae monitor (Ka11'1an 9/ I 0) LbRPSf rae monitor (Kaman 12) Monitor Reactor Building ventilation rad monitor (Kaman 3/4, 5/6, 7 /8) Turbine Building venti l ation rad monitor (Kaman 1/2) Offgas Stack rad monitor (Kaman 9/10) 59 1.0E 01 1:1Ci/ee 1.0E O 1 1:1Ci/ee 4 .5E+02 1:1Ci/ee 1.0E O I 1:1Ci/ee 60 tffil 99 QI (RevisieR
6) tfovemeer 2012 I NE-:1 99 0 I (ReYisioA
6) 1-loYemaer 20 J 2 J __ _.( .... sH-ite~s~pe~c::;.i.i+ifi.f\-C-l'lffi~o=wA:i-+ifi=tOH'r-+I 1-',i Sef-t -a-aARid~tn:w:re,es<.i:11:w:oM<

I dA->.<v-a-al1-11ufee,....s) 2 R 1.3 Dose assessment using act ual meteorology indicates doses greater than 100 mrem TEDE or 500 mrem thyroid CDE at or beyond (site specific dose receptor poiAt)the SITE BOUNDARY. (Preferred] Field survey results indicate EITHER of the fo ll owing at or beyond (site specific dose receptor poiAt)the SITE BOUNDARY:

  • Closed window dose rates greater than 100 mR/h r expected to cont inu e for 60_-minutes or l o n ger.
  • Ana l yses of field s urv ey samp l es indi cate thyroid CDE greater than 500 mrem for one hour of inhalation. 61 62 NEI 99 g l (ReYisioA
6) November 2912 Definitions:

NBI 99 0 I (ReYisioH

6) NoYernber 2012 SITE BOUNDARY:

That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. 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 (PA Gs). 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. This JC is modified by a note that EAL RS 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actual meteorological data and current radiological conditions. However, if Kaman monitor readings are sustained for 15 minutes or longer and the required MIDAS dose assessments cannot be completed within this period, then the declaration can be made using Kaman readings PROVIDED the readings are not from an isolated flow path. 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. If Kaman readings are not valid, field survey results may be utilized to assess this IC using EAL RSJ .3. E scalation of the emergency classification level would be via IC AG+RG 1. Develeper Netes: While this IC may not be met absent challenges to multiple fission product barriers , it provides classification diversity and may be used to classify events that 'Nould not reach the same EGL based on plant status or the fission product matri>c. alone. For many of the DBAs analyzed in the Updated Final Safety Analysis Report, the discriminator will not be the number of fission product barriers challenged , but rather the amount of radioactivity released to the environment. The EPA PAGs are ex.pressed in terms of the sum of the effective dose equivalent (EDE) and the committed effective dose equivalent (CEDE), or as the thyroid committed dose equivalent (CDE). For the purpose of these IC/EALs , the dose quantity total effective dose equivalent (TEDE), as defined in 10 CFR § 20 , is used in lieu of" ... sum of EDE and CEDE .... ". The EPA PAG guidance provides for the use of adult thyroid dose conversion factors; however , some states have decided to base protective actions on child thyroid CDE. Nuclear power 63 Jl.ffil 99 01 (Re*,risieA

6) Nevefl'll:ler 2012 plant ICs/EALs need to be consistent with the protective action methodologies employed by the States within their EPZs. The thyroid CDE dose used in the IC and EALs should be adjusted as necessary to align with State protective action decision making criteria. The " site specific monitor list and threshold values" should be determined with consideration of the following:
  • Selection of the appropriate installed gaseous effluent monitors.
  • The effluent monitor readings should correspond to a dose of 100 mrem TEDE or 500 mrem thyroid CDE at the " site specific dose receptor point" (con s istent with the calculation methodology employed) for one hour of e>,posure.
  • Monitor readings will be calculated using a set of assumed meteorological data or atmospheric dispersion factors; the data or factors selected for use should be the same as those employed to calculate the monitor readings for I Cs AA 1 and AG 1. Acceptable sources of this information include , but are not limited to , the RETS/ODCM and values used in the site's emergency dose assessment methodology.
  • The calculation of monitor readings 1 Nill also require use of an assumed release isotopic mix; the selected mi>, should be the same as that employed to calculate monitor readings for ICs ,6J\l and AGL Acceptable sources of this information include , but are not limited to , the RETS/ODCM and values used in the site's emergency dose assessment methodology.
  • Depending upon the methodology used to calculate the EAL values , there may be overlap of some values between different ICs. DeYelopers will need to address this overlap by adjusting these values in a manner that ensures a logical escalation in the EGL. The " site specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish between on site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan , and the procedural methodology used to determine offsite doses and ProtectiYe Action R e commendations.

The variation in selected dose receptor points means there may be some differences in the distance from the release point to the calculated dose point from site to site. Developers should research radiation monitor design documents or other information sources to ensure that 1) the E AL valu e being considered is v r ithin the usable response and display range of the instrument , and 2) there are no automatic features that may render the monitor reading invalid (e.g., an auto pur g e feature triggered at a particular indication level). It is recognized that the condition described by this IC may result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. In those cases , EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available. For example , an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provi s ion notwithstanding , if the estimated/calculated monitor reading is greater than apprmcimately 110% of th e highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold. Although the IC references TEDE , field survey results are generally available only as a "*whole body" dose rate. For this reason , the field survey EAL specifies a " closed 1 ,¥indow" survey reading. 64 . . Nll! 99 QI (R .. ;,; ** 0) _l*d1eat 1 ees frem a ,ea! time . . """"" 20!2 Many licensees do not ha,'e th* do:: pr0ject1on system are not incl . seepe eftlie plaRI Teeliei;al

" e'7
al,1_hty. Fe, these tl,at de tlie ea:~~*,'" tlie ge*erie EALs. time dese prnj eetie* system , pe~, . eat, ees. A Ii****** may ;.qoest ;;: . , '{ :*y Rel 1,e wit!, ie tlie esu ts , approval v,rill be considered me u e an EAL using real IRdieatieRs frem . 0" a ease hy ease 1,asis. Many licensees do not h ~, pen~eter m?~itoring system are not in I . aRd maietaiRed te tlie s a.e ,tl,,s e"f3ah1hty.

Fe, tl,ese tl,at de ti, e ode<l rn !lie geee,ie BAL,. s . ame eve! as pl 1 . , ese meRito,s . . peeifieatiees. le a<iditiee , . -* --ae-eqo1pmeet , e, witliie the see may eet he eeRt,elled lieeesee may reqoest te i '1 :a<i1Rgs may 1,e iRlloeeeed 1,y OR"ire pe ~ftlie plaet Teelieieal ,eeesidered ee a ease hy ::: : ae EAL osi*g a pe,imete, me~iter~meRta er etlier faete,s. A e as1s. ng system; approval will be 3.1.3.C E~C::,:1Lb-ff~.s'rnsfi1i gtt;ftfm'RelenR1th,Ac\:t1tt:firiH:b~u~te*sr.

~,.i.....::t...c.

65 NEI 99 QI (RevisioA e) No*,ember 2Q 12 AS2RS2 [See Develeper )Vates] ECL: Site Area Emergency Initiating Condition: Spent fuel pool level at (site specific Level 3 description) 16.36 feet. Operating Mode Applicability: All Example Emergency Action Levels: R 2.1 Lowering of spent fuel pool level to 16.36 feet.(site specific Level 3 value). Definitions: Basis: This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration. It is recognized that this IC would lik e ly not be met until well after another Site Area Emergency IC was met; however , it i s included to provide classification diversity. Esca lation of the emergency classification level would be via IC AG+--RG I or AmRG2. Developer Notes: ln accordance 1 Nith the discussion in Section 1.4 , NRG Order EA 12 051 , it is recommended that this IC and EAL be implemented when the enhanced spent fuel pool level instrumentation is a:Yailable for use. The " site specific Level 3 value" is usually that spent fuel pool level v,rhere fuel remains covered and actions to implement make up water addition should no longer be deferred. This site specific level is determined in accordance with NRG Order EA 12 051 and NEI 12 02 , and applicable owner's group guidance. Developers should modif)* the EAL and/or Basis section to reflect any site specific constraints or limitations associated v,rith the design or operation of instrumentation used to determine the Level 3 ¥a-H:le-: EGL Assignment Attributes: 3 .1. 3 .B 66 ECL: General Emergency N"El 99 0 I (RevisioA

6) No 1 rember 2012 AR G1 Initiating Condition:

Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem T E DE or 5 , 000 mrem thyroid CDE. Operating Mode Applicability: All Emergency Action Levels: (1 or 2 or 3) Notes: Example Emergeney Aetien Levels: I

  • The Emergency Director should dec l are the General Emergencyevent promptly upon determining that the applicable time has been exceeded , or will likely be exceeded.
  • If an ongoing release is detected and the re l ease start t i me is unknown , assume that the release duration has exceeded 15 minutesthe specified time limit.
  • If the effluent flow past an effluent monitor is known to have stopped due to actions to iso l ate the release path , then the effluent monitor reading is no l onger va l id for classification purposes.

I

  • The pre-calcu l ated effluent monitor values presented in EAL .L l should_QDJy be used for emergency c l assification assessments until the resu l ts from a dose assessment using actua l meteorology are available. fL l I I Reading on ANY of the follovringTable R-1 effluent radiation monito rs greater than the rea a* h I " GE"£ 15 . t I 1ng s ownco umn or mmu es or anger: J Effluent MeniteF Glassifieatien

+lueshelds M0Rit0F Reaetor BllildiAg \*eAtilatioA rad moAitor EKamaH ;3 1 4, ~1 6, '.7 1 8~ +llrbiAe B1JildiAg \'eAtilatioA rad moAitor A<:::aA~aA !12~ Gffgas 8tael , rad moAitor fK:amaA 9110~ Monitor Reactor Bu il ding ventilation r a d monitor (K aman 3/4, 5/6, 7 /8) Turb i n e Bu il ding ventil at i o n r a d monitor aJ (K a m a n 1/2) V) C1l l.!J l.lE+OO uci/cc l.4E+OO u c i/c c 67 GE I .Oe*OO llGi ,l ee l .Oeai=OO 1JGi l ee 4 .~ea1=0;3 llGi l ee

t-ffil 99 QI (ReYisioR e) }>foyember 2Ql2 Dose assessment using actual meteorology indicates doses greater than 1,000 mrem TEDE or 5 , 000 mrem thyroid CDE at or beyond (site specific dose receptor point)the SITE BOUNDARY. [Preferred]

Field survey results indicate EITHER of the following at or beyond (site specific dose receptor point)the SITE BOUNDARY:

  • Closed window dose rate s greater than 1 , 000 mR/hr expected to continue for 60_ minutes or longer.
  • Analyses of field survey samp l es indicate thyroid CDE greater than 5 , 000 mrem for one hour of inhalation.

68 69 l>!El 99 0 I (ReYisioA

6) l>te\'emeer 2012 Definitions:

NEI 99 0 I (Re~*isioA

6) },/ovemeer 2012 SITE BOUNDARY:

That line beyond which the land is neither owned, nor leased, nor otherwise controlled by the l icensee. Basis: This IC addresses a release of gaseous rad i oactivity that results in projected or actual offsite doses greater than or e qu a l to the EPA Protective Action Guides (PAGs). It in c lud es b ot h monitored and un-monitored releases. Releases of this m agn itud e w ill require implementation of protective actio n s for the public. This IC is modified by a note that EAL RG 1.1 is only assessed for emergency classification until a qualified dose assessor is performing assessments using dose projection software incorporating actua l meteorological data and current radio l ogical conditions. However, if Kaman monitor readings are sustained for 15 minutes or l onger and the required MIDAS dose assessments cannot be completed within this period, then the declaration can be made using Kaman readings PROVIDED the readings are not from an isolated flow path. Ra diolo g i ca l effl u e nt EALs are a l so included to provide a b asis for classifying events and co nd itions that ca nn ot be readily or appro pri ate l y classified on the b as i s of plant co nditi ons a lon e. T h e inclu s i o n of both plant condition and radiological effl u e nt EALs mor e fu ll y addresses the spec trum of possible acc id e nt eve nt s a nd conditions. The TEDE dose i s set at the EPA PA G of 1 , 000 mrem w hil e the 5 , 000 mr em thyroid CDE was esta bli s h ed in consideration of the I :5 rat i o of the EPA PAG for TEDE and thyroid C D E. C l ass ifi cat i on based on efflue nt monitor readings ass um es th at a release path to the e n viro nm e nt i s esta bli shed. _If the effl u ent flow past an effl u e nt monitor i s known to have stoppe d du e to act i ons to isolate the re l ease path, then the effluent m o nit or r ead in g i s n o longer va lid for classification purposes. If Kaman readings are not valid, field survey results may be utilized to assess this IC using EAL RG 1.3. Developer Notes: The effluent ICs/EALs are ineluded to provide a basis for elassifying events that eannot be readily elassified on the basis of plant eonditions alone. The inelusion of both types oflCs/EALs more fully addresses the speetrum of possible events and aeeidents. While this IC may not be met absent challenges to multiple fission product barriers, it provides classification diversity and may be used to elassify events that would not reaeh the same EGL based on plant status or the fission product matrix alone. For many of the DB As analyzed in the Updated Final Safety Analysis Report, the diseriminator will not be the number of fission product barriers challenged , but rather the amount of radioactivity released to the environment. The EPA PAGs are e>rpressed in terms of the sum of the effeetive dose equivalent (EDE) and the committed effective dose equiYalent (CEDE), or as the thyroid committed dose equiYalent (CDE). For the purpose of these IC/EALs , the dose quantity total effeetive dose equivalent (TEDE), as defined in 10 CFR § 20, is used in lieu of " ... sum of EDE and CEDE .... ". 70 1-ffil 99 01 (Re\*isioA

6) NoYemaer 2012 The EPA PAG guidance provides for the use of adult thyroid dose conversion factors; however , some states haYe decided to base protective actions on child thyroid CDE. Nuclear power plant ICs/EALs need to be consistent with the protective action methodologies employed by the States within their EPZs. The thyroid CDE dose used in the IC and EALs should be adjusted as necessary to align with State protective action decision making criteria.

The "site specific monitor list and threshold values" should be determined 1.vith consideration of the follo*n<ing:

  • Selection of the appropriate installed gaseous effluent monitors.
  • The effluent monitor readings should correspond to a dose of 1,000 mrem TEDE or 5,000 mrem thyroid CDE at the " site specific dose receptor point" (consistent 1.vith the calculation methodology employed) for one hour of e>(posure.
  • Monitor readings will be calculated using a set of assumed meteoro l ogical data or atmospheric dispersion factors; the data or factors se l ected for use should be the same as those employed to calculate the monitor readings for ICs AAl and ASl. Acceptable sources of this information include , but are not limited to, the RETS/ODCM and values used in the site's emergency dose assessment methodology.
  • The calculation of monitor readings will also require use of an assumed release isotopic mix; the selected mix should be the same as that employed to calculate monitor readings for !Cs AAl and AS!. Acceptable sources of this information include , but are not limited to, the RETS/ODCM and values used in the site's emergency dose assessment methodology.
  • Depending upon the methodology used to calculate the EAL values, there may be overlap of some Yalues between different ICs. Developers will need to address this overlap by adjusting these values in a manner that ensures a logical escalation in the EGL. The " site specific dose receptor point" is the distance(s) and/or locations used by the licensee to distinguish between on site and offsite doses. The selected distance(s) and/or locations should reflect the content of the emergency plan, and procedural methodology used to determine offsite doses and Protective Action Recommendations.

The yariation in selected dose receptor points means there may be some differences in the distance from the release point to the calculated dose point from site to site. Developers should research radiation monitor design documents or other information sources to ensure that 1) the EAL value being considered is 1.vithin the usable response and display range of the instrument , and 2) there are no automatic features that ma)' render the monitor reading invalid (e.g., an auto purge feature triggered at a particular indication level). It is recognized that the condition described by this IC may result in a radiological effluent value beyond the operating or display range of the installed effluent monitor. In those cases, EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available. For example , an El\L monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than appro>(imately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold. Although the IC references TEDE, field survey results are generally available only as a " whole body" dose rate. For this reason , the field survey EAL specifies a "closed window" surYey reading. 71 NEI 9 9 g I (Re'\'isioA <i) Noyember 2Ql2 . I ded in the generic EALs. . * , tern are not me u

  • h
  • the I time dose pr0Ject1on S) s b" lit , may not be vt'lt m Jnel i eatiens freffi 8 reah. 1,; 1 ity For those that <lo, the eapa . 1 t el 88 B',L *sing real d t ha"e t 1s capa * , Efuest to me u e ' Many lieense~so;o hni;al Sjleeifieations.

A lieensee "'"; red en a ease 1,y ease 1,asis. scope of the pant ec Its* approval ,,viii be cons 1

  • fen systeffi """ '
  • 1,, L tiffie <Iese pr~e,r1 . 1 <lee! in the genef!e,'

5* . erimeter monitoring system are not me u monitors may not be controlled Jnelieat1ens ffeffi 8 P h. al, ility Fer these tliat <lo , these fthe plant Teehnieal . d not have t 1s cap * . ,.,*thin the scope o A Many lieen_seese I, e leYel as plant eq*1pffient, er "'. :.. ffie~al er other faetors: n aoel maiota,neel te t e '""' eliogs "'"Y 1,e i ofl*eneeel

9) en,_,ron *steffl* approval w1ll l,e 8 ecifications. In add1t1_on, rea AL ing a perimeter momtormg Sy , li:ensee may "'q*est te ,nol*<I: !ffi en considered on a case by case as1s. . t Attributes*

3.1.4.C EGL Ass1gnmen n

  • 72 Jl>JBI 99 g l (ReYisioA
6) Novemeer 2Q 12 AG2RG2 [Sec Dc 1;clepa lilotcs] ECL: General Emergency Initiating Condition:

Spent fuel pool level cannot be restored to at least 16.36 feet~ (site specific Level 3 description) for 60_-minutes or longer. Operating Mode Applicability: All Example Emergency Action Levels: Note: The Emergency Director should declare the General Emergency event promptly upon determining that the applicable time 60 minutes has been exceeded , or will likely be exceeded. R Q 2.1 Spent fuel pool level cannot be restored to at least 16.36 feeh (site specific Level 3 value) for 60 minutes or longer. Definitions: 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. It is recognized that thi s IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity. DevelepeF Notes: In accordance with the di s cussion in Section 1.4 , NRG Order EA 12 051 , it is recommended that this IC and EAL be implemented when the enhanced spent fuel pool level instrumentation is available for use. The " site sp e cific L e vel 3 value" is usually that spent fuel pool level where fuel remains covered and actions to implement make up *water addition should no longer be deferred. This site specific level is determined in accordance with : NRG Order EA 12 051 and NEI 12 02 , and applicable owner's group guidance. Developers should modify the EAL and/or Basis section to reflect any site specific constraints or limitations associated with the design or operation of instrumentation used to determine the Level 3 ¥aHie-: E GL Assignment Attributes: 3 .1.4 .G 73 NEI 99 0 I (Re\*i sioA e) November 2012 7 COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS Table C 1: Ree0gniti0n Categerv "C" Initiating C0nditi0n Matrix UNUSUAL EVENT CUl UNPLi\J,J1'ffiD loss of (reactor vesse l/RC£ [PWR] or RPV [BWR]) inventory for 15 minutes or longer. Op. 1\1edes: Geld Shutdewn , Refueli~ fr CU2 Loss of all but one AC power souree to emergency buses for 15 m i nutes or longer. Op. Afedcs: LJCeld S/n,ttde,rn, Re/Meling , Defueled CUJ ill-l"PLA}J1'1ED increase in RC£ temperature. Op. },1edcs: LJCeld Shuldew;'l , Refueling CU4 Loss of Vital DC po*n<er for 15 minutes or longer. Op. },fedcs: LJCeld Sh1itdewn , Refueling CUS Loss of all onsite or offsite communications capabilities. Op. },fedes: LJCeld 8hutdov,rn , Re.fitelir1g , Dafaeled ALERT CAl Loss of (reactor Yessel/RC8 [P WR] or R.0 V [BWR]) inventory. Op. 1\1edc s: LJCeld Shutdewn , Refiwling CA2 Loss of all offsite and all onsite AC power to emergency buses for 15 minutes or longer. Op. },1edcs: LJCeld Shutdewn , Refueling , Defiwled CAJ Inability to maintain the plant in cold shutdown. Op. },1edc s: LJCeld Shutdewn , Refueling 74 8ITEAREA GENERAL EMERGENCY EMERGENCY C81 Loss of (reactor vessel/RC£ [PWR] or fil>V [BWR]) CG 1 Loss of (reactor vessel/RC£ [PWR] or RPV [BWR]) inventory affecting core decay heat removal capability. Op. 1\1edcs: LJCeld Shutdewn , Refueling inventory affecting fuel elad integrity v,ith containment challenged. Op. A1edcs: LJCeld Shi1tde*,im , Refueling ,-------------------, 1 Table intenEieEi for use by 1 I I 1 EAL Se\*elopers. 1 : Inclusion in licensee 1 I S ' , ,l 1 ocuments 1s not requ1reu. 1 L------------------* UNUSUAL EVENT ALERT CA(i Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode. Op. ,\1edcs: LJCeld Shuu/ewn , Refueling 75 8ITEAREA EMERGENCY I NEI 99 QI (ReYisioR e) No,*ember 2() 12 GENERAL EMERGENCY Table intended for use b;' 1 EAL developers.

Inclusion in licensee : documents is not required. L------------------1 NE! 99 QI (ReYisioR
6) 1'Jo1rember 2Q12 CU1 ECL: Notification of Unusual Event Initiating Condition:

UNPLANNED loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory for 15 minute s or lon g er. Operating Mode Applicability: Cold Shutdown , Refueling4, 5 Emergency Action Levels: Example Emergeney Aetien Levels: (1 or 2) Note: The Emergency Director should declare the Unusual Eventevent promptly upon determining that the applicable time 15 minutes has been exceeded , or will lik ely be exceeded. UNPLANNED loss of reactor coolant results .fin (reactor vessel/RCS [PWR] or RPV [BWR]) level less than a required lower limit for 15 minutes or longer. a. (Reactor vessel/RCS [PWR] or RPV_ [BWR]) level cannot be monitored. --AND --+---b. UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool increase in (site specific sump and/or tank) Suppression Pool or Drywell and Reactor Building floor and equipment drain sump levels. Definitions: UNPLANNED: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected p l ant response to a transient. The cause of the parameter change or event may be known or unknown. Basis: This IC addresses the inabi li ty to restore and maintain water level to a required minimum level (or the lower limit of a level band), or a Joss of the ability to monitor (reactor vessel/RCS [PWR] er RPV [BWR]) level concurrent with indications of coo lant 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 l evel decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is avai labl e to keep the core covered. 76 NEI 99 Ql (ReYisioR

6) 1-fo*,remser 2Q 12 EAL CUl.1 recognizes that the minimum required (reactor vessel/RCS

[PWR] or RPV [BWR]) 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. 77 78 *NEI 99 QI (ReYisioR e) ~fo\'emaer 2Q 12 ~JEJ 99 0 I (RevisioR

6) November 2012 ---EAL CUI .2 addresses a condition where all means to determine (reactor vessel/RCS

[PWR] or RPV [BWR]) 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 eYaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor 'vessel/RCS [PWR] or R..1 H/ [BWR]). If all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that RSC inventory loss was occurring by observing sump and Suppression Pool level changes. The drywell floor and equipment drain sumps, reactor building equipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor building. A rise in Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Pool level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA l or CA3. De¥elopeF Notes: EAL #1 It is recognized that the minimum allmvable reactor vessel/RCS/RP\' level may have many values over the course of a refueling outage. Developers should solicit input from licensed operators concerning the optimum wording for this EAL statement. In particular , determine if the generic wording is adequate to ensure accurate and timely classification , or if specific setpoints can be included without making the EAL statement umYieldy or potentially inconsistent with actions that may be taken during an outage. If specific setpoints are included , these should be drawn from applicable operating procedures or other controlling documents. E AL #2.b E nter any " site specific sump and/or tank" levels that could be e>(pected to increase ifthere were a loss of inventory (i.e., the lost inventory would enter the listed sump or tank). EGL Assignment Attributes: 3.1.1.A 79 ECL: Notification of Unusual Event NEI 99 QI (RevisieA

6) ]'Jeyemeer 2012 CU2 Initiating Condition:

Loss of all but one AC power source to emergency essential buses for 15 minutes or longer. Operating Mode Applicability: Cold Shutdown, Refuelingi,2 , Defueled Example Emergency Action Levels: Note: Th e E mergency Director should declare the Unusual Eventevent promptly upon determining that 15 minutes has been exceeded , or will likely be exceeded. cµ2.1 a. AC power capability to (site specific emergency buses) 1 A3 and l A4 buses is reduced to a sin g le power source for 15 minutes or longer. AND b. Any additional single power source failure will result in loss of all-ALL AC power to SAF E TY SYSTEMS. Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition. including the ECCS. These systems are classified as safety-related .A system required for safe plant operation, cooling dov,rn the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems classified as safety related. Basis: This IC de s cribes a significant degradation of offsite and onsite AC power sources such that any additional s ingle failure would result in a loss of all AC power to SAFETY SYSTEMS. ln this condition , the sole AC power source may be powering one , or more than one , train of safetyrelated 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 emer g ency bus. Some examples of this condition are presented below.

  • A loss of all offsite power with a concurrent failure of all but one emergency power source ( e.g., an onsite diesel generator).
  • A loss of all offsite power and loss of all emergency power sources (e.g., onsite diesel generators) with a single train of emergency essential buses being back fed from the unit main generator.
  • A loss of emergency power sources (e.g., onsite diesel generators) with a single train of 80 emergency essential buses be i ng -eae-k-fed from an offsite power source. 81 :t-1:el 99 g 1 (Re¥isi0R e) }fo>,*emaer 2Q 12 NE! 99 Q 1 (Re\*ision e) }fo*reA'!aer 2Ql2 Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power. The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2. 82 NEJ 99 Q 1 (Re\1 isioA 6) 1-Jovemeer

?Q 12 Developer Notes: For a po'1.*er source that has multiple generators, the EAL and/or Basis section should reflect the minimum number of operating generators necessary for that source to provide required po'w*Jer to an AC emergency bus. For example, if a backup po'IJer source is comprised of h.vo generators (i.e., tv.*o 50% capacity generators sized to feed 1 AC emergency bus), the EAL and Basis section must specify that both generators for that source are operating. The "site specific emergency buses" are the buses fed by offsite or emergency AC po'.ver sources that supply po'.*.*er to the electrical distribution system that po'wvers SAFETY SYSTEMS. There is typically 1 emergency bus per train of SAFETY SYSTEMS. Developers should modify the bulleted examples provided in the basis section, above, as needed to reflect their site specific plant designs and capabilities. The EALs and Basis should reflect that each independent offsite po'w\'er circuit constitutes a single po'wver source. For example, three independent 345k'I offsite power circuits (i.e., incoming po'wver lines) comprise three separate po'1Jer sources. Independence may be determined from a review of the site specific UFSAR, 83 ~rn r 9 9 01 (R ev i s ioR 6) Nov em e er 2012 SBO analysis or related loss of electrical po'.ver studies. The EAL and/or Basis section may specify use of a non safety related po 1.ver source provided that operation of this source is recognized in AOPs and EOPS, or beyond design basis accident response guidelines (e.g., FLEX support guidelines). Such po'.ver sources should generally meet the "Alternate ac source" definition provided in 10 CFR 50.2. At multi unit stations, the EALs may credit compensatory measures that are proceduralized and can be implemented 1.vithin 15 minutes. Consider capabilities such as power source cross ties, "sv.*ing" generators, other power sources described in abnormal or emergency operating procedures, etc. P lants that have a proceduralized capability to supply offsite AC po 1.-1er to an affected unit via a cross tie to a companion unit may credit this po'wver source in the EAL pro 1 1ided that the planned cross tie strategy meets the requirements of 10 CFR 50.63. ECL Assignment Attributes: 3.1.1.A 84 ECL: Notification of Unusual Event Initiating Condition: UNPLANNED increase in RCS temperature. Operating Mode Applicability: Cold Shutdown, RefuelingU Emergency Action Levels: NEI 99 0 l (Re*,isioH

6) ~foyemeer 2012 CU3 Example Emergeney Aetien Levels: ( 1 or 2) Note: The Emergency Director should declare the Unusual Eventevent promptly upon determining that the applicable time 15 minutes has been exceeded , or will likely be exceeded.

~1 ~2 UNPLANNED increase in RCS temperature to greater than (site specific Technical Specification cold shutdown temperature limit)212°F. Loss of ALL RCS temperature and (reactor vessel/RCS [PWR] or RPV [BWR]) level indication for 15 minutes or longer. Definitions: 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. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. For DAEC, this is considered to be Secondary Containment as required by Technical Specifications. CONTAil'J:MENT CLOSURE: 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. Basis: 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 CLOSURE is not established during this event , the Emergency Director 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. 85 1'ffil 99 g 1 (ReYisioR e) NoYemeer 2Q 12 EAL CU3.l 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. Refuelin g evolutions that lower water l eve l below the reactor vessel flange are carefully planned and controlled. A lo ss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown. 86 NEI 99 0 I (ReYisioA

6) }fo,*emeer 2012 EAL CU3.2 reflects a condition where there has been a significant loss of in strumentation capab ilit y necessary to monitor RCS conditions and operators wou ld 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 h eat l oad has been reduced since the cessation of power operation. Fifteen minutes was selected as a thresho ld to exclude transient or momentary losses of indication.

Escalation to A l ert would be v i a IC CAI based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time cr i teria. Develof)eF Notes: for EAL #1 , enter the " site specific T e chnical Specification cold shutdovm temperature limit" where indicated. E GL Assignment Attributes: 3.1.1.A 87 ECL: Notification of Unusual Event Initiating Condition: Loss of Vital DC power for 15 minutes or longer. Operating Mode Applicability: Cold Shutdo 1 ,1rn, Refueling1_,_i Example EmergeeeyEmergency Action Levels: NEI 99 01 (ReYisioR

6) No,*ember 2012 CU4 Note: The E mergency Director should declare the Unusual E~vent promptly upon determining that the applicable time 15 minutes has been exceeded , or will likely be exceeded.

C 4.1 Indicated vo lt age is l ess than (site specific bus voltage value)] 05 VDC on BOTH Div 1 and Div 2 125 VDC busesrequired Vital DC buses for 15 minutes or longer. Definitions: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related .A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the EGGS. Systems classified as safety related. 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 availab l e 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 schedu l ed outage maintenance work and Train B is in-service (operable), then a loss of Vital DC power affecting Tra in B would require the declaration of an Unusual Event. A loss of Vita l 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. Depending upon the event , escalation of the emergency c l assification l evel would be via IC CA 1 or CA3 , or an IC in Recognition Category AR. DeYeleper Netes: The " site specific bus voltage value" should be based on the minimum bus voltage necessary for adequate operation of SAFETY SYSTEM equipment. This Yoltage value should incorporate a margin of at least 15 minutes of operation before the onset of inability to operate those loads. This Yoltage is usually near the minimum voltage selected when battery sizing is performed. 88

~ NEI 99 01 (Re11isioR

6) No11ember 2012 The typical value for aA eAtire battery set is apprmdmately 105 VDC. For a 60 cell striAg of batteries , the cell voltage is approximately 1.75 Volts per cell. For a 58 striAg battery set , the miAimum voltage is apprmdmately 1.81 Volts per cell. EGL AssigAmeAt Attributes:

3.1.1.A 8 9 NEl 99 01 (Revisien

6) ~fo\cemeer 2012 ECL: Notification of Unusual Event Initiating Condition:

Loss of all onsite or offsite communications capabilities. Operating Mode Applicability: Cold Shutdown, Refueling5, 64, 5 , Defueled Emergency Action Levels: f ... etien Le1,*els: (1 or 2 or 3) C 5.1 Loss of ALL of the following onsite communication methods: CU5 _* _(site specific list of communications methods)Plant Operations Radio System

  • In-Plant Phone System
  • Plant Paging System (Gaitronics)

C l,J 5.2 Loss of ALL of the following GRGoffsite response organization communications methods:

  • DAEC All-Call phone
  • All telephone lines (PBX and commercial)
  • Cell Phones (including fixed cell phone system)
  • Control Room fixed satellite phone system
  • FTS Phone system (site specific list of communications methods) C 5.3 Loss of ALL of the following NRC communications methods:
  • FTS Phone system
  • All telephone lines (PBX and commercial)
  • Cell Phones (including fixed cell phone system)
  • Control Room fixed satellite phone system * (site specific list of communications methods) 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 GRGoffsite response organization s and the NRC. 90 NEI 99 Ql (ReyisioA

6) ~loYember 2Q 12 This IC should be assessed only when extraordinary means are being utilized to make communications possible ( e.g., use of non-plant , privately owned equipment , relaying of on-site information via individuals or multiple radio transmission points, individuals being sent to offsite locations , etc.). 91

}J:El 99 QI (RevisioA

6) }lo,*emeer

?Q 12 EAL CU5.1 addresses a total l oss of the communications methods used in support of routine plant operations. EAL CU5.2 addresses a total loss of the communications methods used to notify a ll GRGoffsite response organization s of an emergency declaration. The offsite response organizations referred to here are the State of Iowa, Linn County, and Benton CountyThe OROs referred to here are (see Developer Notes). ---EAL CU5.3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration. Developer Notes: EAL #1 The "site specific list of coFRFRunications methods" should include all coFRFRunications FRethods used for routine plant comFRunications (e.g., coFRFRercial or site telephones, page party systeFRs, radios, etc.). This listing should include installed plant equipFRent and coFRponents, and not items owned and FRaintained by individuals. EAL #2 The "site specific list of coFRFRunications FRethods" should include all coFRmunications FRethods used to perforFR initial emergency notifications to OROs as described in the site EFRergency Plan. The listing should include installed plant equipFRent and coFRponents, and not iteFRs owned and FRaintained by individuals. farnFRple FRethods are ring down/dedicated telephone lines, coFRFRercial telephone lines, radios, satellite telephones and internet based coFRFRunications technology. In the Basis section, insert the site specific listing of the OROs requiring notification of an eFRergency declaration froFR the Control RooFR in accordance with the site EFRergency Plan, and typically v,ithin 15 FRinutes. EAL #3 The "site specific I ist of comFRunications FRethods" should include all coFRmunications FRethods used to perforFR initial eFRergency notifications to the NRG as described in the site EFRergency Plan. The listing should include installed plant equipFRent and coFRponents, and not iteFRs ovmed and FRaintained by indiYiduals. These FRethods are typically the dedicated EFRergency Notification System (ENS) telephone line and coFRFRercial telephone lines. EGL Assignment Attributes: 3.1.1.C 92 NEI 99 0 I (ReYisioR e) ~loYemaer

20) 2 CA1 ECL: Alert Initiating Condit ion: Loss of (reactor vessel/RCS

[PWR] or RPV [BWR]) inventory. Operating Mode Applicability: Cold Shutdo'.¥n, Refueling4, 5 Emergency Action Levels: Example EmergeneyEmergenev Aetion LeYels: (1 or 2) Note: The Emergency Director should declare the ~event promptly upon determining that the applicable time 15_ minutes has been exceeded , or will likely be exceeded. Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory as indicated by level less than (site specific level) 1 19 .5 inches. a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored for 15 minutes or longer AND b. UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool UNPLAN}rBD increase in (site specific sump and/or tank) levels due to a loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventor y. Definitions: 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. Basis: This IC addresses conditions that are precursors to a los s of the ability to adequately cool irradiated fuel (i.e., a precursor to a cha ll enge to the fuel c l ad barrier). This cond iti on represents a potential substantial reduction in the level of plant safety. For EAL CAI .1 , a lowerin g of water level below (site specific level) 119.5 inches indicates that operator actions have not been successful in restoring and maintaining (reactor vessel/RCS [PWR] or RPV [BWR]) water l eve l. The heat-up rate of the coolant wi ll increase as the ava il ab l e water inventory is reduced. A continuing decrease in water level will lead to core uncovery. Although related, EAL CA 1 .1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal ( e.g., loss of a Residu a l Heat Removal suction point). An increase in RCS temperature caused by a loss of decay heat remova l capability is eva luat ed under IC CA3. 93 NEI 99 Q 1 (Re't'isioR e) November 2012 For EAL CAl.2 , the inability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) 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 ma)' 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 1.vater flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]).the operators wou l d need to determine that RSGRCS inventory loss was occurring by observing sump and Suppression Pool level changes. The drywe ll floor and equipment drain sumps, reactor building eq u ipment and floor drain sumps receive all liquid waste from floor and equipment drains inside the primary containment and reactor bui l ding. A rise in Suppression Pool water level may be indicative of valve misalignment or leakage in systems that discharge to the Torus. Sump and Suppression Poo l l evel increases must be eva l uated against other potential sources of l eakage such as cooling water sources i nside the containment to ensure they are indicative of RCS leakage. 94 tffil 99 g I (ReYisioR e) tJoyemeer ?QJ2 The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS 1 If the (reactor vessel/RCS [PWR] or RPV [BWR]) inventory level continues to lower , then escalation to Site Area Emergency would be via IC CS 1. Developer Notes: For EAL #1 the " site specific level" should be based on either: * [BWR] Low Low EGGS actuation setpoint/Level

2. This setpoint was chosen because it is a standard operationally significant setpoint at v,hich some (typically high pressure EGGS) injection systems would automatically start and is a value significantly below the low RPV v,ater level RPS actuation setpoint specified in IC GU 1. * [PWR] The minimum allowable level that supports operation of normally used decay heat removal systems (e.g., Residual Heat Removal or Shutdown Cooling).

If multiple levels e>,ist , specify each along with the appropriate mode or configuration dependency criteria. For EAL #2 The type and range of RCS level instrumentation may vary during an outage as the plant moves through various operating modes and refueling evolutions, particularly for a PWR. As appropriate to the plant design, alternate means of determining RCS level are installed to assure that the ability to monitor level v,ithin the range required by operating procedures will not be interrupted. The instrumentation range necessary to support implementation of operating procedures in the Gold Shutdovm and Refueling modes may be different (e.g., narrower) than that required during modes higher than Gold Shutdown. E nter any " site specific sump and/or tank" levels that could be expected to increase if there were a loss of inventory (i.e., the lost inventOF)' would enter the listed sump or tank). EGL Assignment Attributes: 3.1.2.B 95 ECL: Alert 1'ffil 99 QI (Re,*isioR 6) 1'IO'ieFReer 2Q 12 CA2 Initiating Condition: Loss of all offsite and all onsite AC power to emergency essential buses for 15 minutes or longer. Operating Mode Applicability: Cold Shutdown , Refueling4. 5 , Defueled Emergency Act i on Levels: Example EmergeneyEmergeney A.etien Le11els: Note: The Emergency Director should declare the A-left-event promptly upon determining that the applicable time 15_ minutes has been exceeded , or will likely be exceeded. C 2.1 Loss of ALL offsite and ALL onsite AC Power to (site specific emergency buses)1A3 and 1 A4 buses for 15 minutes or longer. Definitions: SAFETY SYSTEM: A system required for safe plant operation. cooling down the plant and/or placing it in the co l d shutdown condition, including the ECCS. These systems are class i fied as safety-related.A system required for safe p l ant operat i on, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. Systems classified as safety related. 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 thresho l d to exclude transient or momentary power losses. ---Escalation of the emergency classification level would be via IC CSl or A8+RSl. Develeper Netes: For a povt1er source that has multiple generators , the EAL and/or Basis section shou l d reflect the minimum number of operating generators necessary for that source to provide adequate power to an AC emergency bus. For e>rnmple , if a backup pm.ver source is comprised of two generators 96 NEI 99 g 1 (RevisioR

6) tlo't'emeer 2912 (i.e., two 50% eapaeity generators sized to feed 1 AC emergeney bus), the EAL and Basis seetion must speeify that both generators for that souree are operating.

The "site specific emergency buses" are the buses fed by offsite or emergency AC pov,er sources that supply power to the eleetrieal distribution system that powers SAFETY SYSTEMS. There is typieally 1 emergency bus per train of SAFETY SYSTEMS. The EAL and/or Basis section may specify use of a non safety related power source provided that operation of this souree is eontrolled in aeeordanee with abnormal or emergeney operating proeedures , or beyond design basis aeeident response guidelines (e.g., FLEX support guidelines). Sueh power sourees should generally meet the " Alternate ae souree" definition provided in 10 CFR 50.2. At multi unit stations , the EALs may credit compensatory measures that are proeeduralized and ean be implemented 1tvithin 15 minutes. Consider eapabilities such as power source cross ties, "s 1 i11ing" generators , other power sources described in abnormal or emergency operating procedures , ete. Plants that have a proeeduralized eapability to supply offsite AC power to an affected unit via a cross tie to a companion unit may credit this power source in the E AL provided that the planned eross tie strategy meets the requirements of 10 CFR 50.63. EGL Assignment Attributes: 3.1.2.B 97 NE! 99 0 I (RevisioA

6) }lo\*emeer 2012 CA3 ECL: Alert Initiating Condition:

Inability to m ai nt a in the plant in cold shutdow n. Operating Mode Applicability: Cold Shutdovm , Refuel in~ Emergency Action Levels: Example Emergeney Aetien Levels: (1 or 2) Note: The E mergenc y Director s hould d e clare th e A-left-event promptly up o n determinin g that the app li ca ble time h as b ee n excee d e d , or w ill lik e l y b e excee d ed. C 3.1 C 3.2 UNPLANNE D increase in R CS temperature to greater than (site specific Technical Specification cold shutdown temperature limit)212°F for grea t e r than the duration spec ifi e d in the following tableTable C-2. Table C-2+ RCS Heat-up Duration Thresholds RCS 8tatuslntegri!Y CONTAINMENT CLOSURE Heat-up Duration Status Intact (but not at reduced Not applicable 60 minutes* in*rentory [PWR]) Not intact (or at reduced Estab li s hed 20 minute s* in*rentor) ' [PWR]) Not Esta blished 0 minute s

  • _If a n RCS h eat removal syste m i s in operation within this time frame and RC S temperature i s being r e duc ed, the EAL i s not applicable.

UNPLANNED RCS pressur e increa se g reater than (site specific pressure reading) IO psig due to a loss of RCS cooling .. (This EAL does not apply during water solid plant conditions. [PWR]) Definitions: UNPLANN E D: A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant respon s e to a transient. The cause of the parameter change or event may be known or unknown. CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures. systems, and components as a functional barrier to fission product release under existing plant conditions. For DA E C, this is considered to be Secondary Containment as required by Technical Specifications. CONTAINMENT CLOSURE: The proc e durally 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. 98 Basis: NEI 99 0 I (Re*,isioR

6) ~JOY8ffl08F 2012 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.

RCS integrity is intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams). 99 NEI 99 QI (Re¥isioA

6) ~lo\*emeer 2Q 12 The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact , or RCS inventory is reduced (e.g., mid loop operation in P'.1/Rs) . .! The 20-min ut e criterion was included to allow time for operator action to address the temperature increase.

The RCS Heat-up Duration Thresho ld s table a l so addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not cruc i a l in this cond iti on since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety. Fina ll y, in the case where there is an increase in RCS temperature, the RCS is not int act or is at reduced inventory [PWR], and CONTAINMENT CLOSURE is not estab li shed , 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 coo lant inv entory above the top of irradiated fuel. EAL CA3.2 provides a pressure-based indication of RCS heat-up. Esca l ation of the emergency c l assification l evel wou ld be via IC CS 1 or AS+RS 1. DeYeloper Notes: For EAL #1 Enter the "site specific Technical Specification cold shutdovm temperature limit" where indicated. The RC8 should be considered intact or not intact in accordance v,ith site specific criteria. For EAL #2 The "site specific pressure reading" should be the lowest change in pressure that can be accurately determined using installed instrumentation , but not less than 10 psig. For PWRs, this IC and its associated EALs address the concerns raised by Generic Letter 88 17, Less ofDee*ly Heat Reme-vel. /\ number of phenomena such as pressuri2:ation, vorte,..ing, steam generator U tube draining , RCS level differences when operating at a mid loop condition, decay heat removal system design , and level instrumentation problems can lead to conditions where decay heat removal is lost and core uncovery can occur. NRG analyses shov,r that there are sequences that can cause core uncovery in 15 to 20 minutes , and severe core damage within an hour after decay heat removal is lost. The allo1,ved time frames are consistent with the guidance provided by Generic Letter 88 17 and believed to be conservative gi 1 1en that a low pressure Containment barrier to fission product release is established. EGL Assignment Attributes: 3.1.2.B 100 ECL: Alert ~ffil 99 01 (RevisioR e) November 2012 CA6 Initiating Condition: Hazardo u s event affect in g a SAFETY SYSTEM n eede d for the current o p erat in g mode. Operating Mode Applicability: Cold Shutdown, Refuelin~ Emergency Action Levels: ExamJJle EmeFgeney Aetien Levels: l. Notes:

  • If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred.

then this emergency classification is not warranted.

  • -If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM. then this emergency classification is not warranted.

C 6.1 a. The occurrence of ANY of the Table C-3 hazardous events:The occurrence of ,'\NY 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 or Emergency Director
  • Seismic event (earthquake)
  • Internal or external flooding event
  • High winds or tornado strike
  • FIRE
  • EXPLOSION (site specific hazards)River level above 757 feet
  • River Water Supply (RWS) pit low level alarm
  • Other events *with similar hazard characteristics as determined by the Shift Manager or Emergency Director ---AND b. EITHER of the following:

101 NEl 99 01 (Re\'isioR 6) }loYeFRber 2012 ------1. Event damage has caused indications of degraded performance in-at least one train of a SAFETY SYSTEM needed for the current operating mode. 2. 2EITHER of the following:.,-

  • Event damage has caused indications of degraded performance to a second train of the SAFETY SYSTEM needed for the current operating mode,-ef, _* _The event has caused resulted in VISIBLE DAMAGE to the second train of a SAFETY SYSTEM component or structure needed for the current operating mode.::;: -. 102 Definitions: "t-ffiI 99 01 (ReYisioR
6) "t-l0\'8fl'!88F 2012 FIRE: Combustion characterized by heat and light. Sources of smoke such a s 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 (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events may require a post-event inspection to determine if the attributes of an e x plosion are present. SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These systems are classified as safety-related .A system required for safe plant operation. cooling down the plant and/or placing it in the cold shutdovm condition. including the EGGS. Systems classified as safety related. 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. Damage resulting from an equipment failure and limited to the failed component (i.e., the failure did not cause damage to a structure or any other equipment) is not VISIBLE DAMAGE. Basis: This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of d e graded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential e x ists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance for criteria CA6. l .b. l of this EAL; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events. An event affecting a single-train SAFETY SYSTEM (i.e., there are indications of degraded performance and/or VISIBLE DAMAGE affecting the one train) would not be classified under CA6 because the two-train impact criteria that underlie the EALs and Bases would not be met. If an event affects a single-train SAFETY SYSTEM, then the emergency classification should be made based on plant parameters/symptoms meeting the EALs for another IC. Depending upon the circumstances, classification may also occur based on Shift Manager/Emergency Director judgement. 103 Nel 99 QI (Re*,isioR e) No¥ember 2() 12 Indications of degraded performance 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. VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. 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. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train. 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 f*ftAh EAL l .b. l 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 sigAificant 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 eyent 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 AS1RS1. De:velepeF Netes: For (site specific hazards), developers should consider including other sigAificant , site specific hazards to the bulleted list contained in EAL l .a (e.g., a seiche). Nuclear po 1 n<er plant SAFETY SYSTEMS are comprised of two or more separate and redundant trains of equipment in accordance 1.vith site specific design criteria. EGL Assignment Attributes: 3.1.2.B 104 NEI 99 QI (Re,,*isioR a) ~lo 1 rem0er 2012 CS1 ECL: Site Area Emergency Initiating Condition: Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory affecting core decay heat removal capability. Operating Mode Applicability: Cold Shutdown , Refuelin~ Emergency Action Levels: Example Emergeney Aetion LeYels: (1 or 2 or3) Note: The Emergency Director should declare the Site Area Emergencyevent promptly upon determining that the applicable time 30 minutes has been exceeded , or will likely be exceeded. ~1 a. b. ~2 a. b. CONTAINMENT CLOSURE not established. AND (Reactor vessel/RCS [PWR] or RPV [BWR]) level LESS THANless than specific level)+64 inches" CONTAINMENT CLOSURE established. AND (Reactor vessel/RCS [PWR] or RPV [BWR]) level LESS THA1'J:less than specific level).+ 15.:.: inches ~3 a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level cannot be monitored for 30 minutes or longer. AND b. Core uncovery is indicated by AN¥-EITHER of the following: I Definitions:

  • (Site.33 ecific radiation monitor) Drywell Monitor (9184A/B) reading greater than (site :33 ecific value)5.0 R/hr
  • Erratic source range monitor indication

[PWR]

  • UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool UNPLANNED increase in (site specific sump and/or tank)_levels of sufficient magnitude to indicate core uncovery * (Other site specific indications) 105 tl-EI 99 QI (ReYisioR
6) tlovemaer 2012 CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

For DAEC, this is considered to be Secondary Containment as required by Technical Specifications. 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. 106 Basis: NE! 99 01 (ReYisioA

6) NoYemeer 2012 This IC addresses a significant and prolonged loss of (reactor vessel/RCS

[PWR] or RPV [BWR]) inventory control and makeup capability leading to IMMINENT fuel dama g e. 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. 107 NET 99 01 (ReYisioA e) November 2012 Following an extended loss of core decay heat removal and inventory makeup , decay heat will cause reactor coolant boilin g and a further reduction in reactor vessel level. If RG-8,l reactor vessel level cannot be restored , fuel damage is probable. Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINM E NT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RG-8,l reactor vessel levels of EALs CS 1.1.b and CS 1.2.b reflect the fact that with CONTAINMENT CLOSURE established , there is a lower probability of a fission product releas e to the environment. -.. In the Cold Shutdown and Refueling Modes, LT/LI-4559 , 4560 , and 4561 (RX VESSEL NARROW RANGE LEV E L) instruments read up to 22" high due to hot calibrations. LI-4541 (WR GEMAC , FLOODUP) should be used in these Modes for comparison to EAL thresholds s ince it is calibrated cold and reads accurately. If normal means of RPV level indication are not available due to plant evolutions, redundant means of RPV level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. In EAL CS 1.3.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. The inabilit y to monitor (reactor vessel/RCS [PWR] or RPV [BW~]) 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 [PWR] or RPV [BWR]). These EALs address concerns raised by Generic Letter 88-17, Lo s s of Deca y Heat Removal; SECY 91-283 , Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Pow e r Operation at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guid e lin es for Industry Actions to Assess Shutdown Management. ---Escalation of the emergency classification level would be via IC CGI or AG+RGl. De, 1 el0f)er Notes: Accident analyses suggest that fuel damage may occur vrithin one hour of uncovery depending upon the amount of time since shutdown; refer to Generic Letter 88 17, SECY 91 283, NUREG 1449 and NUMARC 91 06. The type and range of RCS level instrumentation may vary during an outage as the plant moves through various operating modes and refueling evolutions, particularly for a PWR. As 108 tl-el 99 01 (ReYisioR

6) NoYember 2012 appropriate to the plant design , alternate means of determining RCS level are installed to assure that the ability to monitor level 1 Nithin the range required by operating procedures will not be interrupted.

The instrumentation range necessary to support implementation of operating procedures in the Cold Shutdo\lm and Refueling modes may be different (e.g., narro\1,'er) than that required during modes higher than Cold Shutdovm. For EAL #1.b the " site specific level" is 6" below the bottom ID of the RCS loop. This is the le 1 1el at 6" below the bottom ID of the reactor vessel penetration and not the low point of the loop. If the availability of on scale level indication is such that this leYel value can be determined during some shutd01, 1 ,'n modes or conditions , but not others , then specify the mode dependent and/or configuration states during 1tvhich the level indication is applicable. If the design and operation of water leYel instrumentation is such that this level value cannot be determined at any time during Cold Shutdovm or Refueling modes , then do not include EAL #1 (classification will be accomplished in accordance 1 ,vith EAL #3). For EAL #2.b The "site specific level" should be apprmdmately the top of active fuel. If the availability of on scale le 1 ,rel indication is such that this level value can be determined during some shutdown modes or conditions , but not others, then specify the mode dependent and/or configuration states during 1 ,vhich the level indication is applicable. If the design and operation of water level instrumentation is such that this level value cannot be determined at any time during Cold Shutdovm or Refueling modes , then do not include EAL #2 (classification will be accomplished in accordance with EAL #3). For EAL #3 .b first bullet As water level in the reactor vessel lowers , the dose rate above the core will increase. Enter a " site specific radiation monitor" that could be used to detect core uncovery and the associated " site specific value" indicative of core uncovery. It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases , EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available. For e>rnmple, an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold. To further promote accurate classification , developers should consider if some combination of monitors could be specified in the EAL to bui Id in an appropriate level of corroboration between monitor readings into the classification assessment. For EAL #3.b second bullet Post TMI accident studies indicated that the installed PWR nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations. For EAL #3.b third bullet Enter any 'site specific sump and/or tank" levels that could be e>(pected to change if there were a loss of RCS/reactor vessel inventory of sufficient magnitude to indicate core uncovery. Specific level values may be included if desired. For EAL #3.b fourth bullet Developer s should determine if other reliable indicators exist to identify fuel uncovery (e.g., remote viewing using cameras). The goal is to identify any unique or 109 NEI 99 0 I (Re1t*isioR 6) No;remeer 2012 site specific indications , not already used else*where , that 1tvill promote timely and accurate emergency classification. For EAL #1.b " site specific level" is the LoY,r Low Low EGGS actuation setpoint / Level 1. The BWR Lov,* Lov,r Low EGGS actuation setpoint / Level 1 was chosen because it is a standard operationally significant setpoint at which some (typical!) ' low pre s sure EGGS) injection systems would automatically start and attempt to restore R.0 V level. This is a RPV water level value that is observable below the Lov,r Low/Level 2 value specified in IC CAI , but significantly above the Top of Active Fuel (TOAF) threshold specified in EAL #2. For EAL #2.b The " site specific level" should be for the top of active fuel. For EAL #3 .b first bullet As *;,rater level in the reactor vessel lowers , the dose rate above the core will increase. Enter a " site specific radiation monitor" that could be used to detect core uncovery and the associated " site specific value" indicative of core uncovery. It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases , EAL values should be determined 1.Yith a margin sufficient to ensure that an accurate monitor reading is available. For e>rnmple , an EAL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than approximately 110% of the highest accurate monitor reading , then developers may choose not to include the monitor as an indication and identify an alternate EAL threshold. To further promote accurate classification , developers should consider if some combination of monitors could be specified in the EAL to build in an appropriate level of corroboration between monitor readings into the classification assessment. For BWRs that do not have in s talled radiation monitors capable of indicating core uncovery , alt e rnate site specific level indications of cor e uncovery should b e used if available. For EAL #3 .b second bullet Because BWR source range monitor (SRM) nuclear instrumentation detectors are typically located below core mid plane , this may not be a viable indicator of core uncovery for BWRs. for EAL #3.b third bullet Enter any " site specific sump and/or tank" levels that could be e>,pected to change if there were a loss ofRPV inventory of sufficient magnitude to indicate core uncovery. Specific level values may be included if desired. For EAL #3.b fourth bullet Developers should determine if other reliable indicators e1dst to identify fuel unco 1 ,1ery (e.g., remote viewing using cameras). The goal is to identify any unique or site specific indications , not already used elsewhere , that will promote timely and accurate emergency classification. EGL Assignment Attributes: 3 .1.3 .B 110 J!,ffil 99 g l (ReYisioR

6) NoYemeer 2Q 12 CG1 ECL: General Emergency Initiating Condition:

Loss of (reactor vessel/RCS [PWR] or RPV [BWR]) inventory affecting fuel clad integrity with containment challenged. Operating Mode Applicability: Cold Shutdown , Refueling4. 5 Emergency Action Levels: Examf)le EmergeneyEmergeney Aetion Levels: (1 or 2) Note: The Emergency Director should declare the General Emergencyevent promptly upon determining that the applicable time 30 minutes has been exceeded , or will likely be exceeded. C 1.1 a. (Reactor vessel/RCS [PWR] or RPV [BWR]) level LESS THA1'tless than fs+te-specific level)+ 15 !!inches for 30 minutes or longer. C 1.2 AND bl!. ANY indication from the Secondary Containment Challenge Table (see below)C-1. a. (Reactor ve ss el/RCS [PWR] or RPV [BWR]) level cannot be monitored for 30 minutes or longer. -----AND b. Core uncovery is indicated by EIHERAN¥ of the following:

  • Drywel I Monitor (9184A/B) (Site specific radiation monitor) reading GREATER THANgreater than (site specific value)5.0 R/hr.
  • Erratic source range monitor indication

[PWR]

  • UNPLANNED level rise in Drywell/Reactor Building Equipment or Floor Drain sump, or Suppression Pool illiPLA1'J1'JED increase in (site specific sump and/or tank) levels of sufficient magnitude to indicate core uncovery ,. * (Other site specific indications)

AND 111 Nel 99 Q l (Re*,isioR e) No*,ember 2Q 12 c. ANY indication from the Secondary Containment Challenge Table (see below~ l}. Table C-1 Secondary Containment Challenge Table

  • CONTAINMENT CLOSURE not established*
  • Drywell Hydrogen or Torus Hydrogen GREATER THA}lgreater than 6% AND Drywell Oxygen or Torus Oxygen GREATER THANgreater than 5% (Explosive mixture) exists inside containment
  • UNPLANNED increase in containment pressure
  • Secondary containment radiation monitors above max safe operating limits (MSOL) of EOP 3, Table 6radiation monitor reading above (site specific value) [BWR]
  • If CONTAINMENT CLOSURE is re-established prior to exceeding the 3 0-_minute time limit , then declaration of a General Emergency is not required.

112 Definitions: "t>IEI 99 Q 1 (ReYisioR e) "t>lo1~emeer 2Q 12 CONTAINMENT CLOSURE: Procedurally defined actions taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. For DAEC, this is considered to be Secondary Containment as required by Technical Specifications.CONTAINMENT CLOSURE: The procedurally defiAed 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 be known or unknown. Basis: 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 RGS1 reactor vessel level cannot be restored , fuel damage is probable. With CONTAINMENT CLOSURE not established , there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is 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 burn (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. In EAL CG 1.2.~, 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 113 NEI 99 QI (ReYisioA

6) J>loYeA'leer 2Q 12 for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

For EAL CGl.2.b, the calculated radiation level on the Drywell Monitors (9184A/B) is without the reactor head in place. Calculated in radiation levels with the reactor head in place are below the normal variation in background readings of these monitors. 114 ~ffil 99 01 (Re,*isioR 6) ~lo't'ember 2012 The in ability to monitor (reactor vessel/RCS [PWR] or RPV [BWR]) l eve l may be caused by instrumentation and/or power failures , or water level dropping below the range of ava il able instrumentation. If water level cannot be monitored , operators may determine that an inventory loss i s occurring by observing c h anges in s ump and/or tank l evels. S ump and/or tank l eve l changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the (reactor vessel/RCS [PWR] or RPV [BWR]). F or the Containment Challenge Table, Secondary Containment ma x safe operating (MSOL) limits from EOP 3 are defined as the highest parameter value at which neither: (]) equipment necessary for the safe shutdown of the plant will fail nor (2) personnel access necessary for the safe s hutdown of the plant will be preclud e d. +I h ese 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 Actions to Assess Shutdown Management. Develeper Netes: Accident analyses suggest that fuel damage may occur v,ithin one hour of uncovery depending upon the amount of time since shutdo,*rn; refer to Generic Letter 88 17 , SECY 91 283, NUREG 1449 and NUMARC 91 06. +he type and range of RCS level instrumentation may vary during an outage as the plant moves through various operating modes and refueling evolutions , particularly for a P\VR. As appropriate to the plant design, alternate means of determining RCS level are installed to assure that the ability to monitor level 'Nithin the range required by operating procedures 1 n<ill not be interrupted. +he instrumentation range necessary to support implementation of operating procedures in the Cold Shutdown and Refueling mod e s may be different (e.g., narrower) than that required during modes higher than Cold Shutdown. For EAL #1.a +he " site specific level" should be approximately the top of active fuel. If the availability of on scale level indication is such that this level value can be determined during some shutdown modes or conditions , but not others , then specify the mode dependent and/or configuration states during which the level indication is applicable. If the design and operation of 'Nater level instrumentation is such that this level value cannot be determined at any time during Cold Shutdovm or Refueling modes, then do not include EAL #1 (classification will be accomplished in accordance with EAL #2). For EAL #2.b first bullet As water level in the reactor vessel lowers , the dose rate above the core will increase. Enter a " site specific radiation monitor" that could be used to detect core uncovery and the associated " site specific value" indicative of core uncovery. It is recognized that the condition described by this IC may result in a radiation value beyond the operating or display range of the installed radiation monitor. In those cases , EAL values should be determined with a margin sufficient to ensure that an accurate monitor reading is available. For eJrnmple, an E AL monitor reading might be set at 90% to 95% of the highest accurate monitor reading. This provision notwithstanding , if the estimated/calculated monitor reading is greater than 11 5 ~ffil 99 01 (Re,*isioA 6) NoYeA'laer 2012 apprmdmately 110% of the highest accttrate monitor reading , then developers may choose not to inclttde the monitor as an indication and identify an alternate EAL threshold. To farther promote accttrate classification , developers shottld consider if some combination of monitors cottld be specified in the EAL to bttild in an appropriate level of corroboration between monitor readings into the classification assessment. For BWRs that do not have installed radiation monitors capable of indicating core ttncovery, alternate site specific level indications of core ttncovery shottld be ttsed if available. For EAL #2.b second bttllet Post TMI accident stttdies indicated that the installed PWR nttclear instrnmentation 1.vill operate erratically 1 Nhen the core is ttncovered and that this shottld be ttsed as a tool for making sttch determinations. Becattse BWR Sottrce Range Monitor (SR.M:) nttclear instrnmentation detectors are typically located belov,r core mid plane , this may not be a viable indicator of core ttncovery for BWRs. For EAL #2.b third bttllet Enter any " site specific sttmp and/or tank" levels that cottld be e>(pected to change if there were a loss of inventory of sttfficient magnitttde to indicate core ttncovery. Specific level valttes may be included if desired. For EAL #2.b fottrth bttllet Developers shottld determine if other reliable indicators e>(ist to identify foe! ttncovery (e.g., remote viewing ttsing cameras). The goal is to identif)' any ttniqtte or site specific indications, not already ttsed else 1.vhere , that will promote timely and accttrate emergency classification. For the Containment Challenge Table: Site shtttdown contingency plans typically pro 1 ,ride for re establishing CONTAINMENT CLOSURE following a loss of RCS heat removal or inventory control fonctions. For " E>(plosive mi>(tttre" , de1+*elopers may enter the minimttm containment atmospheric hydrogen concentration necessary to sttpport a hydrogen bttrn (i.e., the lower deflagration limit). A concttrrent containment 0>1.ygen concentration may be inclttded if the plant has this indication available in the Control Room. For BWRs , the use of secondary containment radiation monitors shottld provide indication of increased release that may be indicative of a challenge to secondary containment. The "site specific valtte" shottld be based on the E OP ma>1.imttm safe valttes becattse these valttes are easily recognizable and have a defined basis. EGL Assignment Attribtttes: 3.1.4 .B 116 tlEI 99 QI (ReYisioR

6) }lo~*emeer 2Q 1 2 8 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI) ICS/EALS Table E 1: Reeognition Category "E" Initiating Condition Matrix UNUSUAL EVENT E HUl Damage to a loaded cask CONFINEMENT BOUNDARY.

Op. },ledes: All 117 I Table intenEleEI for use b)' 1 EAL EleYelopers.

lnelusion in lieensee I ,.i * * ,.i 1 uoeurnents 1s not requ1reu.

1 L------------------* I8F8I MALFUNCTION ECL: Notification of Unusual Event Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY. Operating Mode Applicability: All Example Emergency Action Levels: E-HU1 E HUI .1 Dama ge to a loaded cask CONFINEMENT BOUNDARY as indicated by an on contact£! radiation reading greater than the values shown belewon Table E-1 (2 times the site specific cask specific technical specification allowable radiation level) on the surface of #le-spe nt fuel cask. Table E-1 Cask Dose Rates 61BT DSC 3 feet from HSM Surface 800 mrem/hr Outside HSM Door-Centerline of DSC 200 mrem/hr End Shield Wall Exterior 40 mrem/hr Definition: CONFINEMENT BOUNDARY: The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. Basis: This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a sto rage cask containing spent fuel. It applies to irradiated fuel that i s licen se d for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors , and configuration changes which could cause challenges in removing the cask or fuel from storage. The existence of " damage" is determined by radiological survey. The technical specification multiple of " 2 times", which is also used in Recognition Category A-R IC RA Ul , is used here to distinguish between non-emergency and emergency conditions. _ The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the " on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask. Security-related events for ISFSis are covered under !Cs HUI and HAL 118 61[ *0oua!1dwo3 JO 0ll30!J!'J.103 04l u! p0laoo1 UO!ll30!J!Oeds IUO!U400l S ,)fSUO 04i U! pUAOj eq uuo )(SUO !0RJ lU0ds u JOj f0A0! UO!fU!PUJ 01qaN,Ol(l3 04+/- °("Ol0 '3)lfl0HDfV3 '31H.>f 'NOIS0'1dX 3 ')fSl30 J0AO pedd!l JO peddoJp u '*~*e) SlU0Ae epuw uaw JO IUJAll3U p0!J!lU0P! JO e~uaJ 04l WOJJ l(AS0J p1noo ll34l e~awap sesseJppa 'IV3 S!lf+/- "kttVGtffiOff +/-N:3~\i:'3:NHN:03 04l l00JJU ,\l(U!lU0lOd P(AOO lU4l SUO!l!PUOO lU0P!OOl3 pua SlU0lr0 uu0wou04d ft3JAll3U 04l ,\J!lU0P! ':µode"tt UO!ll3A(l3A3 Al0Jl3S 3~N: pell3(0J 04l pua 0oua!1dwo3 JO 0ll30!J!'J.103 )(Sl3o 04l U! p0ou0J0j0J 'ttVS u JO '[9£§ I D~flN: .10d] ("ttVS) :µode~ S!SA(UUV Al0Jl3S JS.>fSI 04uo Sl(AS0J 04+/- N0I.L3NflA'IVJi\t ISASI 9 FISSION PRODUCT BARRIER ICS/EALS Table 9 f l: RecognitioR Category "f" IRitiatiRg CoRditioR Matrix ALERT A Ry Loss or aRy PoteRtial Loss of either the fuel Clad or RCS barrier. Loss or PoteRtial Loss of aRy two barriers. Loss of aRy hNo barriers aRd Loss or PoteRtial Loss of the third barrier. See Table 9 F 2 fer :s,¥R EALs See Table 9 F 3 fer PWR EALs Jl,J:gJ 99 () 1 (Revision e) Jl,lo*remeer 2() 12 Devele~er Nate: The adjaceRt logic tlo\v diagram is for use by deYelopers aRd is Rot required for site specific im):)lemeRtatioR

however , a site specific scheme must iAclude some t)')
)e of user aid to faci I itate timely aRd accurate classificatioA of fissioR product barrier losses aRd/or poteRtial losses. Such aids are typically com):)rised of logic tlov,r diagrams , " scoriRg" criteria or checkbox type matrices.

The user aid logic must be coRsisteRt with that of the adjaceRt diagram. 120 m I I ~I LOSS POTENTI A L LOSS FU E L CLAD LOSS POTENT I AL LOSS FUEL CLAD L OSS POTENT I A L LOSS FU EL CLAD LOSS LOSS LOSS 1/2 POT E NTIAL LOSS POTENTIAL L OSS RCS LOSS POTENTIAL LOSS Y ES [Gl -Lo ss of A N Y Two Barriers A.till. Lo ss or Potentia l Los s of Third Barrier POTENTIAL LOSS CON T A INMENT £SJ. -Loss or Po t ential Loss of ANY Two Barriers ~-------------------- --t,.i .EA.l-ANY Loss or A NY Potential Loss of filill.ER Fuel C lad Q& R CS 121 l'>J"EI 99 0 I (Re'>*isioR 6) 1'1e¥effieer 2012 N N .....; I NE! 99 0 I (Revision

6) NoYemeer 2012 Tab l e 9-FF-14: BWR-DAEC EAL Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FAlALERT ANY Loss OR ANY Potential Loss of EITHER the Fuel Clad OR RCS barrierAey MY.Loss or ~any Potential Loss of either the Fuel Clad or OR RCS barrier. Fuel Cla d Barr i er L O SS POTENTIAL LOSS 1. RCS Activity A. Coolant activity Not Applicable greater than 300 !!Ci/gm dose eguivalent J-filA. (Site specific indications that reactor coolant activity is greater than 300 µCi/gm dose equivalent I +Mt FSl SITE AREA EMERGENCY Loss &rOR Potential Loss of ~ANY two barriers.

RCS Barrier LOSS POTENTIAL LOSS 1. Primary Containment PressureConditions A. Primary Not Applicable containment pressure greater than (site specific ¥a-tt:lej~ due to RCS leakage. 123 FGlGENERALEMERGENCY Loss of ANY two barriers AND Loss OR Potential Loss of the third barrierLoss of any MY_two barriers and Loss or QR.Potential Containment Barrier LOSS POTENTIAL LOSS 1. Primary Containment Conditions A. UNPLANNED A. Primary rapid drop in containmentTorus primary pressure greater containmentDrywe than (site specific ll pressure ¥a-tt:lej 5 3 psi g following primary OR containmentDrywe B. Drywell or Torus 11 pressure rise H2 cannot be OR determined to be B. Primary LeSS +Ht\Nless containmentDrywe than 6% and 11 pressure Drywell &OR response not Torus 02 cannot consistent with be determined to LOCA conditions. be less than 5% OR (site speeifie C. UNISOLABLE explosiYe direct downstream mixture) exists 12athway to the inside primary environment exists eontainment after 12rimary OR Fuel Clad Barrier RCS Barrier LOSS POTE NTIAL LOSS LOSS POTEN TIAL LOSS D. 124 NI;:I 99 0 I (Re*,*isioA 6) N01,cember 2012 Containment Barrier LOSS POTENTIAL LOSS containment C. HC+L (GraRh 4 of isolation signal EOP 2) exceeded. OR Intentional Rrimary containment venting Rer EOPs I Fuel Clad Barrier RCS Barrier I LOSS I POTENTIAL LOSS LOSS POTENTIAL LOSS 2. RPV Water Level 2. RPV Water Level A. SAG entry is A. RPV water level A. RPV water level ot Applicable requiredPrimary cannot be restored cannot be restored eeRtaiRmeRt and maintained and maintained fleediRg required. above Esite speeifie above fs-tte-RJH,l water le*,<el speeifie R::P\l eerrespeRdiRg te water le¥el the tep ef aetive eerrespeRdiRg te .fue.B+ 15 inches OR the tep ef aeti¥e cannot be fue-4+ 15 inches determined. OR cannot be determined.

3. Not Applicable
3. RCS Leak Rate Not Applicable Not Applicable A,_UNISOLABLE A. UNISOLABLE break in ,<\....~¥ ef primary system the follewiRg:

leakage that Esite speeifie results in systems with exceeding the peteRtial fer high Max Normal eRergy liRe OQerating Limit breaks)Main (MNOL) of EOP Steam, HPCI, 3, Table 6 for Feedwater, EITHER of the RWCU, or RCIC following: as indicated by the

  • 1. Max failure of both :ti-formal isolation valves in OperatiRg aRyANY one line Temperature to close AND OR -EITHER:
  • 2. Max
  • High MSL flow or steam Nermal r\ * -A ---'-'r-.......... b --*----125 2. Jl-JEl 99 01 (ReYisioA
6) Jl-JoYeA1ber 20 12 Containment Barrier LOSS POTENTIAL LOSS RPV Water Level Not Applicable A. SAG entry i s requiredPrimary eeRtaiRmeRt fleediRg required. 3. Primary Containment Isolation Failure A. UNISOLABLE Not Applicable Qrimary system leakage that results in exceeding the Max Safe OQerating Limit (MSOL) of EOP 3, Table 6 for EITHER of the following:
  • TemQerature OR
  • Radiation Level UNISO bABbe direet dewR s tream I Fuel Clad Barrier RCS Barrier I LOSS I POTENTIAL LOSS LOSS POTENTIAL LOSS tunnel Radiation tem12erature -be\.<el-Level

~ annunciators OR

  • Direct re12ort of steam release OR -B. Emerge nc y RPV Depr ess urization required. l __ 126 NEJ 99 0 I (RevisioA
6) No,*emeer 20 12 Containment Barrier LOSS POTENTIAL LOSS pati:P,1t 1 ay te tile envirenment e*ists afteF pFimaF~' centainment iselatien signal OR ------B-:

Intentie nal 13FimaF~' centainment ventin g peF OR ----b7 l:l~HSQ LABLE pFimaFJ' system leakage that Fesl:llts in e~rneeEI in g El'.IIIER ef the fullevving: L Ma* Safe QpeFating Tem13erntl:IFe. OR 2. Ma~E Safe Q13eFating AFea R:aEliatien I:,ev-eh I Fuel Clad Barrier RCS Barrier I LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 4. Primary Containment Radiation

4. Primary Containment Radiation A. Dr~ell Monitor Not Applicable A. Dr~ell Monitor Not Applicable (9I84A/B)

(9I84A/B) reading greater reading greater than 2000 R/hr. than 5 R/hr after OR reactor B. Torus Monitor shutdownk (9185A/B) Primary reading greater ceAtaiAmeAt than 200 R/hr raEliatieA meAiter reaEliAg greater thaA (site specific value). 5. Other Indications

5. Other Indications A. Fuel damage Not A1;mlicableA,.

Not A1mlicableA,. Not A1mlicableA,. assessment (site specific as (site specific (site specific as indicates at least applicable) as applicable) applicable) 5% fuel clad damage.fstte-specific as applicable)

6. E m erge n cy Director Judgme n t 6. Emergency Director Judgment A. ANY condition in A. ANY condition in A. ANY condition in A. ANY condition in the opinion of the the opinion of the the opinion of the the opinion of the Emergency Emergency E mergency Emergency Director that Director that Director that Dir ector that indicate s Loss of indicate s Potential indicates Loss of indicates Potential 127 4. NET 99 01 (Re,*isioR 6) No 1 ,ember ?Q 12 Containment Barrier LOSS POTENTIAL LOSS Primary Containment Radiation Not Applicable A. Dr~ell Monitor (9184A/B) reading greater than 5000 R/hr. OR B. Torus Monitor (9185A/B) reading greater than 500 R/hrA,. Primary ceAtaiAmeAt raEliatieA meAiter reaEliAg greater thaA (site specific 11*ah,1e). 5. Other Indications Not A1mlicableA,.

A. Fuel damage (site specific as assessment applicable) ~ASAP +.~j indicates at least 20% fuel clad damage.fstte-specific as applicable)

6. Emergency Director Judgment A. ANY condition in A. ANY condition in the opinion of the the opinion of the Emergency Emergency Director that Director that indicates Loss of indicates Potential Fuel Clad Barrier RCS Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS the Fuel Clad Loss of the Fuel the RCS Barrier. Loss of the RCS Barrier. Clad Barrier. Barrier. 128 NEJ 99 0 I (RevisioR
6) November 2012 Containment Barrier LOSS POTENTIAL LOSS the Containment Loss of the Barrier. Containment Barrier.

Basis Information For BWR-DAEC EAL Fission Product Barrier Table 9-FF-1-i NET 99 0 I (Revision

6) 1'Jo,*ember 2012 BWR-DAEC FUEL CLAD BARRIER THRESHOLDS:

The Fuel Clad barrier consists of the zircalloy or stainless steel fuel bundle tubes that contain the fuel pellets. 1. RCS Activity Loss I.A This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm dose equivalent J-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred , it represents a loss of the Fuel Clad Barrier. It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications. There is no Potential Loss threshold associated with RCS Activity. Develeper Netes: Threshold values should be detennined assuming RCS radioactivity concentration equals 300 µCi/gm dose equivalent T 131. Other site specific units may be used (e.g., µCi/cc). Depending upon site specific capabilities , this threshold may have a sample analysis component and/or a radiation monitor reading component. Add this paragraph (or similar wording) to the Basis if the threshold includes a sample analysis component , " lt is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless , a sample related threshold is included as a backup to other indications." 2. RPV Water Level 130 Loss 2.A The Loss threshold represents any EOP requirement for entry into the Severe Accident Guidelines. NEI 99 01 (Re\*isioR e) No,*ember 201? This is identified in the BWROG EPGs/SAGs when adequate core cooling cannot be assured.The Loss threshold represents the EOP requirement for primary containment flooding. This is identified in the BWROG EPGs/8AGs ,vhen the phrase , " Primary Containment flooding Is Required ," a-ppears. Since a site specific RPV water level is not specified here , the Loss threshold phrase , " PrimaF)' containment flooding required ," also accommodates the EOP need to flood the primary containment v,hen RPV water level cannot be determined and core damage due to inadequate core cooling is believed to be occurring. Potential Loss 2.A This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling. 131 BWR FUEL CLAD BARRIER THRESHOLDS: J!>JEI 99 o I (ReYisioR a) J!>Je,,eR1ber 2012 The RPV water level threshold is the same as RCS barrier Loss threshold 2.A. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency. 132 DAEC FUEL CLAD BARRIER THRESHOLDS {cont.): NEI 99 0 I (ReYisioR

6) }>fo,*ember 2012 This threshold is considered to be exceeded when , as specified in the site speeifie EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually , automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization.

EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed I imits. EOPs also specify depressurization of the RPV.in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of avai I able injection sources. Therefore , this Fuel Clad barrier Potential Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted , giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available , precluding RPV depressurization in an attempt to minimize loss of RPV inventory. 133

  • I I I I I I DAEC FUEL CLAD BARRIER THRESHOLDS (eent.)::

Nel 99 O I (RevisioH a) }Jovember 2012 The term " cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. In high-power A TWS/failure to scram events , EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier , the immediate need to reduce reactor power is the higher priority. For such events , ICs ~SA6 or SS~§_ will dictate the need for emergency classification. Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier , a potential loss of the fuel clad barrier is specified. 134 BWR FUEL CLAD BARRIER THRESHOLDS: DeYeloper Notes: Loss 2.A NEI 99 0 I (RevisioA

6) No 1 ,eA~eer 2012 The phrase , " Primary containment flooding required ," should be modified to agree with the site specific EOP phrase indicating e>(it from all EOPs and entry to the SAGs (e.g., dry,Nell flooding required , etc.). Potential Loss 2.A The decision that "RPV water level cannot be determined" is directed by guidance given in the RPV \vater level control sections of the EOPs. 3. Not Applicable (included for numbering consistency between barrier tables) 135 DAEC FUEL CLAD BARRIER THRESHOLDS

{cont.): 4. Primary Containment Radiation Loss 4.A and Loss 4.B Nel 99 O I (RevisioA

6) November 2012 The Drywell and Torus radiation monitor reading 2 correspond s to an instantaneous release of all reactor coolant mass into the Drywell or primary Toruscontainment , assuming that reactor coolant activity equals 300 µCi/gm dose equivalent I-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred , it represents a loss of the Fuel Clad Barrier. The radiation monitor readin~ in this threshold t&-are higher than that specified for RCS Barrier Loss threshold 4.A since it indicates a loss of both the Fuel C l ad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency. There is no Potential Loss threshold associated with Primary Containment Radiation. 136 B'WR DA.EC FUEL CLAD BARRIER THRESHOLDS (eoet.): NEI 99 () 1 (Re\*isioR
6) l>levember

?() 12 her Indications-..

5. 1. Other Iedieatioes Loss and/or Potential Loss 5.A Results obtained from procedure PASAP 7.2, Fuel Damage Assessment, indicate at least 5% fuel clad damage. ffhis subcategory addresses other site specific thresholds that may be included to indicate loss 2:!.:QQ!ential loss of the Fuel Clad barrier based o~la~ecific design characteristics not considered in the generic guidance.

There is no Potential Loss threshold associated with Other Indications. Develof)er Notes: Loss and/or Potential Loss 5 .A Developers should determine if other reliable indicators exist to evaluate the status of this fission product barrier (e.g., reviev, accident analyses described in the site Final Safety Analysis Report, as updated). The goal is to identify any unique or site specific indications that \Viii promote timely and accurate assessment of barrier status. Any added thresholds should represent appro>£imately the same relative threat to the barrier as the other thresholds in this column. Basis infonnation for the other thresholds may be used to gauge the relative barrier threat level. &6_:...._Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad Barrier is lost. Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Fuel Clad Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentiall y lost in the event that barrier status cannot be monitored. 137 00 M ...... BWR--DAEC RCS BARRIER THRESHOLDS: NEI 99 01 (RevisioR

6) No,*ember 2012 The RCS Barrier is the reactor coolant system pressure boundary and include s the RPV and all reactor coolant system piping up to and including the isolation valves. 1. Primary Containment PressureConditions Loss l.A The (site specific value) primary containment~

pressure is the drywell high pressure scram setpoint which indicates a LOCA by automatically initiating the-ECCS or equivalent makeup system . .:. There is no Potential Loss threshold associated with Primary Containment Pressure. DeYel0per Notes: 2. RPV Water Level Loss 2.A This ,vater I+ 15 inches e;iel-corresponds to the top of active fuel (T AF) and is used in the E OPs to indicate challenge to core cooling. The RPV water level threshold is the same as Fuel Clad barrier Potential Loss threshold 2.A. Thus , this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency. This threshold is considered to be exceeded when , as specified in the site specific EOPs , RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually , automatically or by failure of the RCS barrier) or when procedural guidance or a lack of lo w pressure RPV injection sources preclude Emergency RPV depressurization EOPs allow the operator a w ide choice of RPV injection sources to consider when restoring RPV water le vel to within prescribed limit s. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with lo w-pressure injection sources. In some events , elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore , this RCS barrier Loss is met only after either: 1) the RPV has been depres s urized , or required emergency RPV depressurization has been attempted , giving the operator an opportunity to assess the capability of l ow-pressure inject ion sources to restore RPV water level or 2) no low pressure RPV injection systems are available , precluding RPV depressurization in an attempt to minimize loss of RPV inventory. 139 B\¥R DAEC RCS BARRIER THRESHOLDS: "f)>ffiJ 99 0 I (RevisioA e) No 1 1emeer 2012 The term, " cannot be restored and maintained above ," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel , but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained. 140 DAEC RCS BARRIER THRESHOLDS (cont.): NET 99 QI (Re,*ision 6) l>Joven~ser 2Q 12 In high-power ATWS/failure to scram events , EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCR WL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events , ICs SAS or SSS will dictate the need for emer g ency classification. There is no RCS Potential Loss threshold associated with RPV Water Level. 3. RCS Leak Rate Loss Threshold 3.A Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated from the Control Room , the RCS barrier Loss threshold is met. Loss Threshold 3.B Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed , the plant operators are directed to open safety relief valves (SRYs) and keep them open. Even though the RCS is being vented into the suppression pool , a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary. Potential Loss Threshold 3.A Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC , HPCI , etc., which indicate a direct path from the RCS to areas outside primary containment. A Max Normal Operating Limit (MNOL) value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly. 141 L B\\'R DAEC RCS BARRIER THRESHOLDS: NEI 99 0 I (Re,,*ision 6) November 2012 The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes , valves , and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the sy s tem. An UNISOLABLE leak which is indicated by Max Normal OperatingMNOL values escalates to a Site Area E mergency when combined with Containment Barrier Loss threshold 3.A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded. DAEC RCS BARRIER THRESHOLDS {cont.): Developer Notes: Loss Threshold 3.A The list of systems included in this threshold should be the high energy lines which , if ruptured and remain unisolated , can rapidly depressurize the RPV. These lines are typically isolated by actuation of the Leak Detection system. Large high energy line breaks such as Main Steam Line (MSL), High Pressure Coolant Injection (HPCI), feedv,rater , Reactor Water Cleanup (RWCU), Isolation Condenser (JC) or Reactor Core Isolation Cooling (RCIC) that are UNTSOLABLE r e present a significant loss of the RCS barrier. 4. Primary Containment Radiation Loss 4.A The Drywellradiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment , assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold 4.A since it indicates a loss of the RCS Barrier only. There is no Potential Loss threshold associated with Primary Containment Radiation. Developer Notes: The reading should be determined assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory , with RCS actiYity at Technical Specification allowable limits , into the primary containment atmosphere. Using RCS activity at Technical Spec i fication allowable limits aligns this threshold with TC SU3. Also , RCS activity at this level will typically result in primary containment 142

B"WR RCS BARRIER THRESHOLDS: NEl 99 01 (RevisioA

6) J>Jo 1 ,eA10er 2012 In some cases, the site specific physical location and sensitivity of the primary containment radiation monitor(s) may be such that radiation from a cloud of released RCS gases cannot be distinguished from radiation emanating from piping and components containing elevated reactor coolant activity. If so , refer to the Developer Guidance for Loss/Potential Loss 5.A and determine if an alternate indication is available.
5. Other Indications There are no Loss or Potential Loss thresholds associated with Other Indications.

DeYeloper Notes: Loss and/or Potential Loss 5 .A Developers should determine if other reliable indicators e>list to evaluate the status of this fission product barrier (e.g., reYiev,* accident analyses described in the site final Safety Analysis Report , as updated). The goal is to identify any unique or site specific indications that 1tvill promote timely and accurate assessment of barrier status. Any added thresholds should represent apprmC:imately the same relative threat to the barrier as the other thresholds in this column. Basis information for the other thresholds may be used to gauge the relative barrier threat level. 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the RCS barrier is lost. Potential Loss 6.A This thresho l d addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The E mergency Director should also consider whether or not to declare the barrier potentiall y lost in the event that barrier status cannot be monitored. Deyeloper Notes: 144 RWR DAEC CONTAINMENT BARRIER THRESHOLDS: t'IBJ 99 O I (Re\*ision e) November 2012 The Primary Containment Barrier includes the drywell , the wetwell , their respective interconnecting paths , and other connections up to and including the outermost containment isolation valves. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.

1. Primary Containment Conditions Loss l .A and l .B Rapid UNPLANNED loss of primary eontainment drywell pressure (i.e., not attributable to drywell spray or condensation effects) following an initial pressure increase indicates a loss of primary eontainmentdrywell integrity.

Primary containmentDrywell pressure should increase as a result of mass and energy release into the primary containment from a LOCA. Thus , primary containmentdrywell pressure not increasing under these conditions indicates a Joss of primary containment integrity. These thresholds rely on operator recognition of an unexpected response for the condition and therefore a specific value is not a s signed. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition. Loss l.C The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS). The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. Jn addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream. Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiologica l releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category RI Cs. 145 DAEC CONTAINMENT BARRIER THRESHOLDS: Loss l.D NEI 99 01 (Revision e) No*,ember 20 I 2 EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed. Intentional venting of primary containment for primary containment pressure or combustible gas control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition. DAEC CONTAINMENT BARRIER THRESHOLDS (cont.): DAEC CONTAINMENT BARRIER THRESHOLDS: Potential Loss l .A The threshold pressure is the primary containmentTorus internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and , thus , represent a Potential Loss of the Containment barrier. Potential Loss l .B If hydrogen concentration reaches or exceeds the lower flammability limit , as defined in plant EOPs , in an oxygen rich environment , a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment , loss of the Containment barrier could occur. Potential Loss l .C The Heat Capacity Temperature Limit (HC+L) is the highest suppression pool temperature from which Emergency RPV Depressurization will not raise:

  • Suppression chamber temperature above the maximum temperature capability of the suppression chamber and equipment within the suppression chamber which may be required to operate when the RPV is pressurized , OR 146 BWR CONTAINMENT BARRIER THRESHOLDS:

J!>J:El 99 Q 1 (Re., 1 isioA 6) J!>Jovemaer 2Q 12 -Suppression chamber pressure above Primary Containment Pressure Limit A , while the rate of energy transfer from the RPV to the containment is greater than the capacity of the containment vent.

  • The HC+L is a function of RPV pressure , suppression pool temperature and suppression pool water level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant and therefore , the inability to maintain plant parameters below the limit constitutes a potential loss of containment.

147 DAEC CONTAINMENT BARRIER THRESHOLDS {cont.):t DeYelof)er Notes: Potential Loss l .B NET 99 QI (Re*,ision

6) J!>l o,*emeer 2Q 12 BWR EPG s/SAGs s pecifically define the limit s a s sociated with e~£plosive mixtures in terms ofdeflagration concentrations of hydrogen and oxygen. For Mk YIT containments the deflagration limits are "6% hydrogen and 5% m£ygen in the drywell or suppression chamber". For Mk IJJ containments , the limit is the " Hydrogen Deflagration Overpressure Limit". The threshold term " explosive mixture" i s synonymous with the EPG/SAG " deflagration limits". Potential Loss l .C Since the HCTL is defined assuming a range of suppression pool water levels as low as the elevation of the downcomer openings in Mk J/11 containments , or 2 feet above the elevation of the horizontal vents in a Mk JII containment , it is unnecessary to consider separate Containment barrier Loss or :Potential Loss thresholds for abnormal suppression pool *Nater level conditions. If desired , developers may include a separate Containment Potential Loss threshold based on the inability to maintain suppression pool v,*ater leYel aboYe the downcomer openings in Mk I/Tl containments , or 2 feet above the ele*,ation of the horizontal vents in a Mk III containment

\Vith R..°V pressure above the minimum decay heat removal pressure , if it will simplify the a s sessment of the suppression pool level component of the HCTL. 2. RPV Water Level There is no Loss threshold associated with RPV Water Level. Potential Loss 2.A The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold 2.A. The Potential Loss requirement for Primary Containment Flooding indicates adequate core cooling cannot be restored and maintained and that core damage is possible. BWR EPGs/SAGs specify the conditions that require primary containment flooding. When primary containment flooding is required , the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to restqre and maintain adequate core cooling. 148 BWR CONTAINMENT BARRIER THRESHOLDS: NEI 99 0 I (RevisioA

6) November 20 1 2 PRA studies indicate that the condition of this Potential Loss threshold could be a core melt sequence which , if not corrected , could lead to RPV failure and increased potential for primary containment failure. In conjunction with the RPV water level Loss thresholds in the Fuel Clad and RCS barrier columns , this threshold results in the declaration of a General Emergency.

DeyelepeF Netes: The phrase , " Primary coAtaiAmeAt floodiAg required ," should be modified to agree *.vith the site specific EOP phrase iAdicatiAg e>dt from all EOPs aAd eAtry to the SAGs (e.g., drywell floodiAg required , etc.). 3. Primary Containment Isolation Failure These thresholds address incomplete containment isolation that allows an UNJSOLABLE direct release to the environment. Loss 3.A The use of the modifier " direct" iA defiAiAg the release path discrimiAates agaiAst release paths through iAterfaciAg liquid systems or miAor release pathways , such a s iAstrumeAt liAes , Aot protected by the Primary CoAtaiAmeAt JsolatioA System (PCIS). The e>ListeAce of a filter is Aot coAsidered iA the threshold assessmeAt. Filters do Aot remove fissioA product Roble gases. IA additioA , a filter could become iAeffective due to iodiAe aAd/or particulate loadiAg beyoAd desigA limits (i.e., reteAtioA ability has beeA e>rneeded) or water saturatioA from steam/high humidity iA the release stream. FollowiAg the leakage of RCS mass iAto primary coAtaiAmeAt aAd a rise iA primary coAtaiAmeAt pressure , there may be miAor radiological releases associated 'Nith allovt1able primary coAtaiAmeAt leakage through various peAetratioAs or system compoAeAts. MiAor releases may also occur if a primary contaiAmeAt isolatioA valve(s) fails to close but the primar)' coAtaiAmeAt atmosphere escapes to aA eAclosed system. These releases do Aot coAstitute a loss or poteAtial loss of primary coAtaiAmeAt but should be evaluated usiAg the RecogAitioA Category A ICs. Loss 3.B EOPs may direct primary contaiAmeAt isolatioA valve logic(s) to be iAteAtioAally b)*passed , eveA if offsite radioactivity release rate limits will be eKceeded. Under these coAditioAs with a valid primary coAtaiAmeAt isolatioA signal , the coAtaiAmeAt should also be coAsidered lost if primary coAtaiAmeAt veAtiAg is actually performed. JAteAtioAal ventiAg of primary coAtaimneAt for primary coAtaiAmeAt pressure or combustible gas coAtrol to the secoAdar)' coAtaiAmeAt aAd/or the enviroAmeAt is a Loss of the CoAtaiAmeAt. VeAtiAg for primary coAtaiAmeAt pressure coAtrol wheA Aot iA aA accideAt situatioA (e.g., to coAtrol pressure below the dr)"well high pressure scram setpoiAt) does Aot meet the threshold coAditioA. 149 Loss 3.GA NEr 99 QI (RevisioR a) 1l-Jo 1 , 1 ember 2012 The Max Safe Operating Limit (MSOL) for Temperature and the Ma>c. Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail , nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required. BWR CONTA.INMENT BARRIER THRE8HOLD8: The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes , valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system. In combination with RCS potential loss 3.A this threshold would result in a Site Area Emergency. There is no Potential Loss threshold associated with Primary Containment Isolation failureRCS Leak Rate. 150 DAEC CONTAINMENT BARRIER THRESHOLDS {cont.):~ DeYeloper Notes: Loss 3.B NET 99 Q 1 (ReYisioA

6) ~Jo,,ember 2Q 12 Consideration may be given to speeifying the speeifie proeedural step within the Primary Containment Control EOP that defines intentional venting of the Primary Containment regardless of offsite radioaetivity release rate. 4. Primary Containment Radiation There is no Loss threshold associated with Primary Containment Radiation.

Potential Loss 4.A The dryw e ll radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary eontainmentdrywell , assuming that 20% of the fuel cladding has failed. The radiation monitor reading for the torus corresponds to an instantaneous release of all reactor coolant mass directly into the torus, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds. NUREG-1228 , Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents , indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring off site protective actions. For this condition to exist, there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency. DeYeloper Notes: NUREG 1228 , Seurce Estimatiens Duringfncide,~t Respense te Se*rere Nuclear Pewer Pf.GmtAccidcnts , provides the basis for using the 20% fuel el adding failure value. Unless there is a site speeifie analysis justifying a different value , the reading should be determined assuming the instantaneous release and dispersal of the reaetor eoolant noble gas and iodine inventory assoeiated with 20% fuel elad failure into the primary eontainment atmosphere. BWR CONTAINMENT Bz" .. RRIER THRESHOLDS:

5. Other Indications There is no Loss threshold associated with Other Indications Loss and/or Potential Loss 5.A 151 J!IJ:EI 99 () 1 (Re1,*isioA
6) J!llo 1 ,em0er 2012 Results obtained from procedure PASAP 7.2, Fuel Damage Assessment, indicate at least 25% fuel clad damage. This subeategory addresses other site specifie thresholds that may be iRcluded to iRdicate loss or poteRtial loss of the CoRtaiRmeRt barrier based OR plaRt specific desigR eharaeteristies Rot eoRsidered iR the geRerie guidaRee.

PASAP 7.2 only shows whether fuel damage is greater than or Jess than 25%, thus this indication is not likely to be declared before containment barrier potential loss 4.A which indicates 20% fuel damage. However, this potential loss threshold adds an additional layer of diversity to the scheme. 152 De,;eleper Notes: Loss and/or Potential Loss 5.A NEI 99 0 I (RevisioA

6) ~Jovember 2012 Developers should determine if other reliable indicators e1(ist to evaluate the status of this fission product barrier (e.g., review accident analyses described in the site Final Safety Analysis Report, as updated).

The goal is to identify any unique or site specific indications that will promote timely and accurate assessment of barrier status. Any added thresholds should represent approJ(imately the same relative threat to the barrier as the other thresholds in this column. Basis information for the other thresholds may be used to gauge the relafr,re barrier threat level. 6. Emergency Director Judgment Loss 6.A This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Containment barrier is lost. Potential Loss 6.A This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored. 153

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6) NoveA10er 2012 The Containment Barrier includes the containment building and connections up to and including the outermost containment isolation valves. This barrier also includes the ma i n steam , feedwater , and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve. Containment Barrier thresholds are used as criteria for escalation of the EGL from Alert to a Site Area Emergency or a General Emergency.

RCS or SC Tube Lealrnge Loss I .A This threshold addresses a leaking or RUPTURED Steam Generator (SG) that is also FAULTED outside of containment. The condition of the SG , 1.vhether leaking or RUPTURED , is determined in accordance \Vith the thresholds for RCS Barrier Potential Loss I .A and Loss I .A , respectively. This condition represents a bypass of the containment barrier. FAULTED is a defined term within the NEI 99 01 methodology; this determination is not necessarily dependent upon entry into , or diagnostic steps within , an EOP. For example , if the pressure in a steam generator is decreasing uncontrollably fp*1rt &jthe FAULTED de.finitif:m] and the faulted steam generator isolation procedure is not entered because EOP user rules are dictating implementation of another procedure to address a higher priority condition , the steam generator is st il l considered FAULTED for emergency classification purposes. The FAULTED criterion establishes an appropriate lower bound on the size of a steam release that may require an emergency classification. Steam releases of this size are readily observable with normal Control Room indications. The lov,er bound for this aspect of the containment barrier is analogous to the lower bound criteria specified in IC SU3 for the fuel clad barrier (i.e., RCS activity values) and IC SU4 for the RCS barrier (i.e., RCS leak rate values). This threshold also applies to prolonged steam releases necessitated by operational considerations such as the forced steaming of a leaking or RUPTURED steam generator directly to atmosphere to cooldown the plant , or to drive an am,iliary' (emergency) feed 1.vater pump. These types of conditions 1.vill result in a significant and sustained release of radioactive steam to the environment (and are thus similar to a FAULTED condition). The inability to isolate the steam flov, without an adverse effect on plant cooldovm meets the intent of a loss of containment. Stearn releases associated with the expected operation of a SG power operated relief valve or safety relief valve do not meet the intent of this threshold. Such releases may occur intermittently for a short period ohirne following a reactor trip as operators process through emergency operating procedures to bring the plant to a stable condition and prepare to initiate a plant cooldown. Stearn releases associated with the unexpected operation of a valve (e.g., a stuck open safety valve) do meet this threshold. 169

PWR CONTAl}f}.4.ENT BAR.~..IER Thresholds: lAadequate Heat Removal There is no Loss threshold assosiated with IRadequate Heat RemoYal. Potential Loss 2.A N~J 99 Ql (Re*,*ision 6) Jl,Joyenleer 2Q 12 This sondition represents an IMMINENT sore melt sequense whish , if not sorrested , sou Id lead to vessel failure and an insreased potential for oontainment failure. For this sondition to ossur, there must already have been a loss of the RC8 Barrier and the Fuel Clad Barrier. If implementation of a prosedure(s) to restore adequate sore sooling is not effestive (sussessful) within 15 minutes , it is assumed that the event trajestory will likely lead to sore melting and a subsequent shallenge of the Containment Barrier. The restoration prosedure is sonsidered "effestive" if sore exit thermosouple readings are desreasing and/or if reastor vessel level is insreasing. Whether or not the prosedure(s) \Viii be effestive should be apparent within 15 minutes. The Emergensy Direstor should essalate the emergensy slassifisation level as soon as it is determined that the prosedure(s) will not be effestive. Severe assident analyses (e.g., NUREG 1150) have sonoluded that funstion restoration prosedures san arrest sore degradation in a signifisant frastion of sore damage ssenarios , and that the likelihood of sontainment failure is \1 ery small in these events. Given this , it is appropriate to provide 15 minutes beyond the required entr)' point to determine if prosedural astions can reverse the sore melt sequence. Developer Notes: 8ome site spesifis EOPs andtor EOP user guidelines may establish desision making criteria sonserning the number or other attributes of thermosouple readings nesessary to drive astions (e.g., 5 CETs reading greater than 1 , 200oF is required before transitioning to an inadequate sore cooling procedure). To maintain consistency v,ith EOPs , these decision making sriteria may be used in the sore exit thermosouple reading thresholds. Potential Loss 2.A. l Enter site specific sriteria requiring entry into a core sooling restoration prosedure or prompt implementation of core cooling restoration astions. A reading of 1 , 200oF on the CETs may also be used. for plants that have implemented Westinghouse Owners Group Emergency Response Guidelines , enter the parameters and values used in the Core Cooling Red Path. 172 (ll V'~ SSO'J SSOdAij JO , (l!.IB0lUf lU0WU!UlUO:) 04l OlU! 0JRj!OJ pup !0Rj %OZ: 4l!A\ P0l0l90SSU A.IOlU0AUI 0UI 01 uu sug 0 '0J04dsmUlU lU0WU!t3lUOO 04l gU!llmssu p0u!WJ0l0P 0q p11104s gu!po0J 04l '0n10A lU0JeJJ*P u ~: * .r snr , ( 1qou lUOjooo Jopo0J 04l JO jt3SJ0ds!~ put3 0St30j0J sno0uelUUlSU! %OZ: 04l gU!SR JOj S!SOq 04l S0P!AOJd 'SlU0p1oov lUOjd J0A\Ocf ~00jOR:N:~~l0. ~;. :l ':s:uo~~~o0ds 0l!S O ~! 0J04l SS0jUfl 01110A 0JRj!OJ gU!PP019 j0Rj . a °M lU0P!OU} U!JRQ SUO!lOW!lS3: 09JROS '8cc [ 03:110:N. V't SSO'J j0!lU0l0cf

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