ML15141A052
| ML15141A052 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 05/07/2015 |
| From: | Capps S D Duke Energy Carolinas |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML15141A047 | List: |
| References | |
| MNS-15-018 | |
| Download: ML15141A052 (135) | |
Text
ATTACHMENT 1EAL BasesCategory:
Subcategory:
H -Hazards and Other Conditions Affecting Plant Safety4 -FireInitiating Condition:
FIRE potentially degrading the level of safety of the plantEAL:HU4.3 Unusual EventA FIRE within the plant PROTECTED AREA not extinguished within 60 min. of the initialreport, alarm or indication (Note 1)Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Mode Applicability:
AllDefinition(s):
FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.
PROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled.
The Protected Area refers to the designated security area around the processbuildings and is depicted in MNS UFSAR Figure 2-4 Plot Plan and Site Area.MNS Basis:NoneNEI 99-01 Basis:This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.EAL-#4T-he intnent of thA 1I mirnute dumrtinn nQ to qwizA thp FIRR ;;Rd tp d6izrrimin~tn
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.. ... .*ala-rmG, othor indicatiRns of a FIREcould be a drop iR firo main automatic of a suppression system, otc.Upon recoeipt, operators will tako prompt actions to confirmn tho yalidity of an initial fire alarm-,indication, Or repot. FoGr E.AL as.esment puFpss, the emrgoGncy declaration clck starts atthe time that. theoinitial alarmn, inictin orrport war, received, and- no-t t-ho timo that aDocument No. Rev. 0 I Page 138 of 272 ATTACHMENT 1EAL Bases~j E51.J. ~j ~ ~**----------~
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~**~45 555, SS *5L4h15 I *J5 I t.r-AL-#IThis E.AL addr-essos receipt of a single fie alarm~, and the eXiStenco of a FIRE is not '.orfedk-- , , v ----i! wav it *V I mm owS m muq, to *5 aiaIIIlm, wrS% , = MyS ,I4,ilt, v urtv a V., toprom.pt actions ..- to. rnfirm the validity of a .-ncilo fire alarm. For EAL ..a.sesmet pupo .e. .the 30 mini t9 Gtnck staftt at the t 6me that the~ initial ;;lnrm War,~ FeG9ived a!nd Rat thA timnA th~it:: i ihkacu -ant varifleiinntL irtinn Wa'R nanrfnrmd A single firo alarm, absent other niains of a FIRE, may be ind-icative of equipment failure-or a s uriu acatlvation, and not an actual FIRE. ForF this resnddoitioal time Is allowed toverf,'thevaidity of the alarm. Tho 30 minute period is a reasonWab a Oun oftime todetrmie f a atua FRE exis-ts-;
how~eveFo, after that time, and absent ifraonto thecOntra, i that an actal FIRE i p rif an actua! FIRE is Verified by a report from the field, then E-AL- #1 irs imdaeyapplicable-,
and the emergency muswt be decared if the FIRE is Rot extinguished w..ithin 15 fminutesA of the thn Ihr "lnrii t11 hVn du ni nmn ilinn
,-.- -- ---this verificati4on occurs Within 30mntsoftec itof the alarm then this; E=ALi not-a -L .0-- II -__JL_ J~D~iIC~DIO iiu no omoroozc ociawai Is arrnmIn addition to a FIRE addressed by EAL #1-HU4.1 or EAL #2HU4.2, a FIRE within the plantPROTECTED AREA not extinguished within 60-minutes may also potentially degrade the levelof plant safety. This basis e,,ends to a FIRE occurrn.g Within the PROTECTED AREA of anISF=SI located outside the plant PROTECTED AREA. [Sentense for plants with an 1SES--.-n All M.................I.
if a FIRE within the plant or ISFSI [for plants with an ISF-SI eutside the plant Protected ArealPROTECTED AREA is of suffticint size to require a response by an offeite firefighting agency(e.g., a locsal town Fire Department),
then the level of plant safet is potontially degraded.
Thedispatch of an off-site firofighting agency to the site requie an emegency doclaratien only if iti s needed to actively support firefighting efforts because the fire is, beyond the capability of theFire Brigade to exinguish.
Declaration is Rot necessary if the agency resources are placed Onstand by, Or supp~oRtig post extinguishment recoer;F Or investigation actions.Basi Relte Dp .*n an* -rmentr. finm AppeRdkn F%Appendix R to 10 CFR 50, states in part:Criterion 3- of Appendix A to this part that systems ,andcomRponents important to safety shall be designed and located to minimize, consistent with other safet requirements, the probability and effect of fires and explo-sion-is."
Document No. IRev. 0 1Page 139 of 272 ATTACHMENT 1EAL Bases^h, gthe offots of firo, thero systems, assoe.ated with ahieViRng mnaintaining 6afo shutdoWn--4 codtosa6Sumo m~ajor impo~tanco to safot bocausdamage to thrm can lead to coro damage resulting from loss o-f c-ola"nt thrQugh boil off no fire many affect safe shutdown systemsand becal use the loss of fu nctio"n ofsystemRs u-ed t mitigate the con.oq. unces.f dersign basis- accidev'ntUs' pogt firfGOndatiOs does not per so imnpact public safety',
the Reood to limit fire damage tosystems required-to- ac-hieve and mnaintain safo shutdown coanditions ors greater than theReed to limit fire damnaqe to those system reuio toq mitigate the conseQUenceS OfIIOsIgqn asIf accIEoIntS.
In addition, Appen~dix R to 10 CFR 50, requires,,
amRong other coensiderations,,
the use of I -hourfire ba -o fo,-r tho encSlosur Of able and equipmenRt asoco, non afot','
ofone Fed undant train (G.2.c).
As usod- in E6AL1 #2, the 30 mninutes to verify a single ala~m is wollwithin this worFt case 1 -hour time period.Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.MNS Basis Reference(s):
- 1. NEI 99-01 HU4I Document No. I Rev. 0 Page 140 of 2721 ATTACHMENT 1EAL BasesCategory:
Subcategory:
H -Hazards and Other Conditions Affecting Plant Safety4 -FireInitiating Condition:
FIRE potentially degrading the level of safety of the plantEAL:HU4.4 Unusual EventA FIRE within the plant PROTECTED AREA that requires firefighting support by an offsitefire response agency to extinguish Mode Applicability:
AllDefinition(s):
FIRE -Combustion characterized by heat and light. Sources of smoke such as slipping drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.
PROTECTED AREA -An area encompassed by physical barriers and to which access iscontrolled.
The Protected Area refers to the designated security area around the processbuildings and is depicted in MNS UFSAR Figure 2-4 Plot Plan and Site Area.MNS Basis:NoneNEI 99-01 Basis:This IC addresses the magnitude and extent of FIRES that may be indicative of a potential degradation of the level of safety of the plant.-~~~~~- .k -it-Ar ~ I t I fl1ýZ la, aw-I .. --- --rtb o ;W.., mai eftti.gIGe ke. ., smeldeHII;I wa~t LUIphU basket).
in additionI
-alarms, othor indications of a FIRE could be a drop in fire main preSSUre, automatic activation of a suppression system, etc.Upon rocoipt, operators Wi11 take prompt actions to G.Gnfirm the Yalidity of an initial fire alarmA,inRdicaton, Or report. For EMAL assoessment
- purposes, the emergency declaration clock starts atthe time that the initia! alarm, indication, Or report was received, and not the timne that arsubsequent verification action was pe~fbrmcd.
Similarly, the fire duration cleck also starts atthe time oreipt of the initial a!armn, in~dication or report-.Document No. IRev. 0 Page 141 of 27ý2 ATTACHMENT 1EAL BasesEAL4-#Th;is EAL addFrol o rFeceipt of a single fire alarm, the oXistRnco of a FIRE Or, not ; ',r;fi;d(i.e., provod or disproved) within 30 mninutos, of tho alarm. UJpon receipt, operatorS will tkthtq 20 mimrda afani ,.ataL. at'14 t~he time that the initial alarm was~ re aived, andr MCA+ th#4 flimaa subseauont
'.rificatien action wa6 norfbrmed-.
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A single fire alarm, abroent othGrFdcains of a FIRE, may be indicWative of equipment f-ailureor sprius acteyafien, and not an actual FIRE. ForF this, reason, additional tkme is allowed toVorf, th ldity of the alarm. The 30 minuto period is, a roasonablo
-amou nt of tame todotormine of an actual FIRE exists; hoe'oVor, afteArtih at- time, and abseent ifrmto to thecontrary, it is assumed that an ac-tual FIRE Is in progressif an actual FIRE is verified by a report fromR the field, the-n EWL V1 isimdaeyapplicable, and the emergency must be dec~lared-if the FIRE is. not eXtinguis6h~d_
within 15-minutes of theqreport. if the alarm is verified to be due to an eqipet failure OrF prosaciain nthmis verifica tion occurs Within 30- minutes of ter cit f the alarm, then this RALIs= oI I and no emergency aec'arai'on is warr3ntep.
EAL--#In ad-l-ition to a FIRE addressed by EAL #1 or EAI #2, a FIRE within the PROTECTED AREA not w., ithin 60 minutes may also p-teRtally degrade the level of pantsafety. Thi6 basis extends to a F4REF ecuri~ng within the PROTECGTE-D AREA of an lSF-Slplant Protocted Ar~ealIf a FIRE within the plant orISFSI [for plantr with an ISESI outside the plant P,-teted Af.alPROTECTED AREA is of sufficient size to require a response by an offsite firefighting agency(e.g., a local town Fire Department),
then the level of plant safety is potentially degraded.
Thedispatch of an offsite firefighting agency to the site requires an emergency declaration only if itis needed to actively support firefighting efforts because the fire is beyond the capability of theFire Brigade to extinguish.
Declaration is not necessary if the agency resources are placed onstand-by, or supporting post-extinguishment recovery or investigation actions.Bass e*oOlated*
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I Document No. Rev. 0 1 Page 142 of 2721 ATTACHMENT 1EAL Bases..;nnh nnn8 .k ,ra ,n~d8iC nhe 9#e rl Of anne nth .£ 9 nvrte nearen-.
na e withn a n$,sf., anqRdamage to therm can o ore damane r4, ultino from IonS of coolant throUah boil off0a a F*.ra ... "- a a 4. n "~ ..*.4,AJ.
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.--Ij ............conaUMOnS 909s not po so 4pc PUNIIC 6afety, the nReed to iimit Tire damage tosystems6 required to achievo n m-aintani~n safle A-hu--t~down condition is grator than thene-e-d to lmtfire damage to those systems requirod to mnitigate the conrsequence odesign basis accid9nts.
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A.m-.M -..5W FetSD. e A * -C IPU 11M IPOREP.Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or SA9.MNS Basis Reference(s):
- 1. NEI 99-01 HU4Document No. Rev. 0 Page 143 of 272 ATTACHMENT 1EAL BasesCategory:
Subcategory:
Initiating Condition:
H -Hazards and Other Conditions Affecting Plant Safety5 -Hazardous GasesGaseous release IMPEDING access to equipment necessary fornormal plant operations, cooldown or shutdownEAL:HA5.1 AlertRelease of a toxic, corrosive, asphyxiant or flammable gas into any Table H-2 rooms orareasANDEntry into the room or area is prohibited or IMPEDED (Note 5)Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, thenno emergency classification is warranted Table H-2 Safe Operation
& Shutdown Rooms/Areas Bldg. Elevation Unit I Room/Area Unit 2 Room/Area ModesAuxiliary 716' P/C, RHole, near 12N1-185, ABPC thru CAD Door, FF59 4Outside CAD 212800 (1 EMXA) 820 (2EM(XA) 3, 4Auxiliary 750'803 (1 ETA) 805 (2ETA) 3, 4702 (Elec. Pene.) 713 (Elec. Pene.) 3Auxiliary 733' 722 (1 EMXB-1) 724 (2EMXB-1) 3, 4705 (1ETB) 716 (2ETB) 3,4Mode Applicability:
3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
IMPEDE(D)
-Personnel access to a room or area is hindered to an extent that extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
MNS Basis:If the equipment in the listed room or area was already inoperable, or out-of-service, beforethe event occurred, then no emergency should be declared since the event will have noadverse impact beyond that already allowed by Technical Specifications at the time of theevent.Document No. Rev. 0 Page 144 of 272 ATTACHMENT 1EAL BasesThe list of plant rooms or areas with entry-related mode applicability identified specify thoserooms or areas that contain equipment which require a manual/local action as specified inoperating procedures used for normal plant operation, cooldown and shutdown.
Rooms orareas in which actions of a contingent or emergency nature would be performed (e.g., anaction to address an off-normal or emergency condition such as emergency
- repairs, corrective measures or emergency operations) are not included.
In addition, the list specifies the plantmode(s) during which entry would be required for each room or area (ref. 1).NEI 99-01 Basis:This IC addresses an event involving a release of a hazardous gas that precludes or impedesaccess to equipment necessary to maintain normal plant operation, or required for a normalplant cooldown and shutdown.
This condition represents an actual or potential substantial degradation of the level of safety of the plant.An Alert declaration is warranted if entry into the affected room/area is, or may be,procedurally required during the plant operating mode in effect at the time of the gaseousrelease.
The emergency classification is not contingent upon whether entry is actuallynecessary at the time of the release.Evaluation of the IC and EAL does not require atmospheric sampling; it only requires theEmorgency Emergency Coordinator's judgment that the gas concentration in theaffected room/area is sufficient to preclude or significantly impede procedurally requiredaccess. This judgment may be based on a variety of factors including an existing job hazardanalysis, report of ill effects on personnel, advice from a subject matter expert or operating experience with the same or similar hazards.
Access should be considered as impeded ifextraordinary measures are necessary to facilitate entry of personnel into the affectedroom/area (e.g., requiring use of protective equipment, such as SCBAs, that is not routinely employed).
An emergency declaration is not warranted if any of the following conditions apply:" The plant is in an operating mode different than the mode specified for the affectedroom/area (i.e., entry is not required during the operating mode in effect at the time of thegaseous release).
For example, the plant is in Mode 1 when the gaseous release occurs,and the procedures used for normal operation, cooldown and shutdown do not requireentry into the affected room until Mode 4." The gas release is a planned activity that includes compensatory measures which addressthe temporary inaccessibility of a room or area (e.g., fire suppression system testing).
" The action for which room/area entry is required is of an administrative or record keepingnature (e.g., normal rounds or routine inspections).
" The access control measures are of a conservative or precautionary nature, and would notactually prevent or impede a required action.* If the equipment in the listed room or area was already inoperable, or out-of-service, beforethe event occurred, then no emergency should be declared since the event will have noDocument No. Rev. 0 Page 145 ATTACHMENT 1EAL Basesadverse impact beyond that already allowed by Technical Specifications at the time of theevent.An asphyxiant is a gas capable of reducing the level of oxygen in the body to dangerous levels. Most commonly, asphyxiants work by merely displacing air in an enclosed environment.
This reduces the concentration of oxygen below the normal level of around 19%, which canlead to breathing difficulties, unconsciousness or even death.This EAL does not apply to firefighting activities that automatically or manually activate a firesuppression system in an area, or to intontional inoFrtig of '"
t.. (1BWR G,').Escalation of the emergency classification level would be via Recognition Category AR, C or FICs.MNS Basis Reference(s):
- 1. Attachment 3 Safe Operation
& Shutdown Areas Tables R-2 & H-2 Bases2. NEI 99-01 HA5IDocument No. I Rev. 0 1 Page 146 of 272 ATTACHMENT 1EAL BasesCategory:
Subcategory:
H -Hazards and Other Conditions Affecting Plant Safety6 -Control Room Evacuation Initiating Condition:
Control Room evacuation resulting in transfer of plant control toalternate locations EAL:HA6.1 AlertAn event has resulted in plant control being transferred from the Control Room to theAuxiliary Shutdown Panels or Standby Shutdown Facility (SSF)Mode Applicability:
AllDefinition(s):
NoneMNS Basis:The Shift Manager (SM) determines if the Control Room is inoperable and requiresevacuation.
Control Room inhabitability may be caused by fire, dense smoke, noxious fumes,bomb threat in or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2).Inability to establish plant control from outside the Control Room escalates this event to a SiteArea Emergency per EAL HS6.1.NEI 99-01 Basis:This IC addresses an evacuation of the Control Room that results in transfer of plant control toalternate locations outside the Control Room. The loss of the ability to control the plant fromthe Control Room is considered to be a potential substantial degradation in the level of plantsafety.Following a Control Room evacuation, control of the plant will be transferred to alternate shutdown locations.
The necessity to control a plant shutdown from outside the ControlRoom, in addition to responding to the event that required the evacuation of the Control Room,will present challenges to plant operators and other on-shift personnel.
Activation of the EROand emergency response facilities will assist in responding to these challenges.
Escalation of the emergency classification level would be via IC HS6.I Document No. Rev. 0 Page 147 of 272 ATTACHMENT 1EAL BasesMNS Basis Reference(s):
- 1. AP/1 (2)/A/5500/17 Loss of Control Room2. MCS-1465.00-00-0022 Appendix R Safe Shutdown Analysis2. NEI 99-01 HA6I Document No. I Rev. 0 1 Page 148 of 272 ATTACHMENT 1EAL BasesCategory:
Subcategory:
H -Hazards and Other Conditions Affecting Plant Safety6- Control Room Evacuation Initiating Condition:
Inability to control a key safety function from outside the Control RoomEAL:HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to theAuxiliary Shutdown Panels or Standby Shutdown Facility (SSF)ANDControl of any of the following key safety functions is not reestablished within 15 min.(Note 1):" Reactivity
" Core Cooling* NCS heat removalNote 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Mode Applicability:
AllDefinition(s):
NoneMNS Basis:The Shift Manager determines if the Control Room is inoperable and requires evacuation.
Control Room inhabitability may be caused by fire, dense smoke, noxious fumes, bomb threatin or adjacent to the Control Room, or other life threatening conditions (Ref. 1, 2).NEI 99-01 Basis:This IC addresses an evacuation of the Control Room that results in transfer of plant control toalternate locations, and the control of a key safety function cannot be reestablished in a timelymanner. The failure to gain control of a key safety function following a transfer of plant controlto alternate locations is a precursor to a challenge to one or more fission product barrierswithin a relatively short period of time.The determination of whether or not "control" is established at the remote safe shutdownlocation(s) is based on Emergency DIieetGt-Coordinator judgment.
The Emergency DiretetDocument No. Rev. 0 Page 149 of 2721 ATTACHMENT 1EAL BasesCoordinator is expected to make a reasonable, informed judgment within (the site .pecific tim.feo-transfeir1 5 minutes whether or not the operating staff has control of key safety functions from the remote safe shutdown location(s).
Escalation of the emergency classification level would be via IC FG1 or CG1MNS Basis Reference(s):
- 1. AP/1 (2)/A/5500/17 Loss of Control Room2. MCS-1465.00-00-0022 Appendix R Safe Shutdown Analysis3. NEI 99-01 HS6Document No. Rev. 0 Page 150 of 272 ATTACHMENT 1EAL BasesCategory:
Subcategory:
H -Hazards and Other Conditions Affecting Plant Safety7 -Emergency Coordinator JudgmentInitiating Condition:
Other conditions existing that in the judgment of the Emergency Coordinator warrant declaration of a UEEAL:HU7.1 Unusual EventOther conditions exist which in the judgment of the Emergency Coordinator/EOF Directorindicate 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 facilityprotection has been initiated.
No releases of radioactive material requiring offsiteresponse or monitoring are expected unless further degradation of SAFETY SYSTEMSoccurs.Mode Applicability:
AllDefinition(s):
SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 1 OCFR50.2):
Those structures, systems and components that are relied upon to remain functional duringand following design basis events to assure:(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which could resultin potential offsite exposures.
MNS Basis:The Emergency Coordinator/EOF Director are the designated onsite individuals having theresponsibility and authority for implementing the MNS Emergency Response Plan. TheOperations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator/EOF Director and takes actions as outlined in the Emergency Plan implementing procedures.
Ifrequired by the emergency classification or if deemed appropriate by the Emergency Coordinator/EOF
- Director, emergency response personnel are notified and instructed to reportto their emergency response locations.
In this manner, the individual usually in charge ofactivities in the Control Room is responsible for initiating the necessary emergency
- response, Document No. I Rev. 0 Page 151 of 272]
ATTACHMENT 1EAL Basesbut Plant Management is expected to manage the emergency response as soon as available to do so in anticipation of the possible wide-ranging responsibilities associated with managinga major emergency (ref. 1).NEI 99-01 Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the Emergency Di-eGter-Coordinator/EOF Director to fall under the emergency classification level description for an NGUE-Unusual Event.MNS Basis Reference(s):
- 1. MNS Emergency Plan section B On-Site Emergency Organization Section B.2 Emergency Coordinator
- 2. NEI 99-01 HU7Document No. Rev. 0 Page 152 of 272]
ATTACHMENT 1EAL BasesCategory:
Subcategory:
H -Hazards and Other Conditions Affecting Plant Safety7 -Emergency Coordinator JudgmentInitiating Condition:
Other conditions exist that in the judgment of the Emergency Coordinator/EOF Director warrant declaration of an AlertEAL:HA7.1 AlertOther conditions exist which, in the judgment of the Emergency Coordinator/EOF
- Director, indicate that 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 involvesprobable life threatening risk to site personnel or damage to site equipment because ofHOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPAProtective Action Guideline exposure levels.Mode Applicability:
AllDefinition(s):
HOSTILE ACTION -An act toward MNS or its personnel that includes the use of violent forceto destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. Thisincludes attack by air, land, or water using guns, explosives, projectiles,
- vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded.
Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on MNS. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).MNS Basis:The Emergency Coordinator/EOF Director are the designated onsite individuals having theresponsibility and authority for implementing the MNS Emergency Response Plan. TheOperations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator andtakes actions as outlined in the Emergency Plan implementing procedures.
If required by theemergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency responselocations.
In this manner, the individual usually in charge of activities in the Control Room isresponsible for initiating the necessary emergency
- response, but Plant Management isexpected to manage the emergency response as soon as available to do so in anticipation ofthe possible wide-ranging responsibilities associated with managing a major emergency (ref.1).Document No. Rev. 0 Page 153 of 272 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the Emergency Dieote-Coordinator/EOF Director to fall under the emergency classification level description for an Alert.MNS Basis Reference(s):
- 1. MNS Emergency Plan section B On-Site Emergency Organization Section B.2 Emergency Coordinator
- 2. NEI 99-01 HA7Document No. Rev. 0 Page 154 of 272 ATTACHMENT 1EAL BasesCategory:
Subcategory:
H -Hazards and Other Conditions Affecting Plant Safety7 -Emergency Coordinator JudgmentInitiating Condition:
Other conditions existing that in the judgment of the Emergency Coordinator/EOF Director warrant declaration of a Site AreaEmergency EAL:HS7.1 Site Area Emergency Other conditions exist which in the judgment of the Emergency Coordinator/EOF Directorindicate that events are in progress or have occurred which involve actual or likely majorfailures of plant functions needed for protection of the public or HOSTILE ACTION thatresults in intentional damage or malicious acts, (1) toward site personnel or equipment thatcould lead to the likely failure of or, (2) that prevent effective access to equipment neededfor the protection of the public. Any releases are not expected to result in exposure levelswhich exceed EPA Protective Action Guideline exposure levels beyond the SITEBOUNDARY.
Mode Applicability:
AllDefinition(s):
HOSTILE ACTION -An act toward MNS or its personnel that includes the use of violent forceto destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. Thisincludes attack by air, land, or water using guns, explosives, projectiles,
- vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded.
Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on MNS. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area)SITE BOUNDARY-Area as depicted in MNS-SLC-16.11.1 Figure 16.11.1-1 SiteBoundary/Exclusion Area BoundaryMNS Basis:The Emergency Coordinator/EOF Director are the designated onsite individuals having theresponsibility and authority for implementing the MNS Emergency Response Plan. TheOperations Shift Manager (SM) initially acts in the capacity of the Emergency Coordinator andtakes actions as outlined in the Emergency Plan implementing procedures.
If required by theemergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency responselocations.
In this manner, the individual usually in charge of activities in the Control Room isDocument No. I Rev. 0 1 Page 155 of 272 ATTACHMENT 1EAL Basesresponsible for initiating the necessary emergency
- response, but Plant Management isexpected to manage the emergency response as soon as available to do so in anticipation ofthe possible wide-ranging responsibilities associated with managing a major emergency (ref.1).NEI 99-01 Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the Emergency Qke~tef-Coordinator/EOF Director to fall under the emergency classification level description for a Site Area Emergency.
MNS Basis Reference(s):
- 1. MNS Emergency Plan section B On-Site Emergency Organization Section B.2 Emergency Coordinator
- 2. NEI 99-01 HS7I Document No. I Rev. 0 Page 156 of 272 ATTACHMENT 1EAL BasesCategory:
Subcategory:
Initiating Condition:
EAL:H -Hazards and Other Conditions Affecting Plant Safety7 -Emergency Coordinator JudgmentOther conditions exist which in the judgment of the Emergency Coordinator/EOF Director warrant declaration of a General Emergency HG7.1 General Emergency Other conditions exist which in the judgment of the Emergency Coordinator/EOF Directorindicate that events are in progress or have occurred which involve actual or IMMINENTsubstantial core degradation or melting with potential for loss of containment integrity orHOSTILE ACTION that results in an actual loss of physical control of the facility.
Releasescan be reasonably expected to exceed EPA Protective Action Guideline exposure levelsoffsite for more than the immediate site areaMode Applicability:
AllDefinition(s):
HOSTILE ACTION -An act toward MNS or its personnel that includes the use of violent forceto destroy equipment, take hostages, and/or intimidate the licensee to achieve an end. Thisincludes attack by air, land, or water using guns, explosives, projectiles,
- vehicles, or otherdevices used to deliver destructive force. Other acts that satisfy the overall intent may beincluded.
Hostile action should not be construed to include acts of civil disobedience orfelonious acts that are not part of a concerted attack on MNS. Non-terrorism-based EALsshould be used to address such activities (i.e., this may include violent acts betweenindividuals in the owner controlled area).IMMINENT-The trajectory of events or conditions is such that an EAL will be met within arelatively short period of time regardless of mitigation or corrective actions.MNS Basis:The Emergency Coordinator/EOF Director are the designated onsite individuals having theresponsibility and authority for implementing the MNS Emergency Response Plan. TheOperations Shift Manager(SM) initially acts in the capacity of the Emergency Coordinator andtakes actions as outlined in the Emergency Plan implementing procedures.
If required by theemergency classification or if deemed appropriate by the Emergency Coordinator, emergency response personnel are notified and instructed to report to their emergency responselocations.
In this manner, the individual usually in charge of activities in the Control Room isresponsible for initiating the necessary emergency
- response, but Plant Management isexpected to manage the emergency response as soon as available to do so in anticipation ofthe possible wide-ranging responsibilities associated with managing a major emergency (ref.1).[Document No. Rev. 0 Page 157 of 272]
ATTACHMENT 1EAL BasesReleases can reasonably be expected to exceed EPA PAG plume exposure levels outside theSite Boundary.
NEI 99-01 Basis:This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrantdeclaration of an emergency because conditions exist which are believed by the Emergency D9ietG-Coordinator/EOF Director to fall under the emergency classification level description for a General Emergency.
MNS Basis Reference(s):
- 1. MNS Emergency Plan section B On-Site Emergency Organization Section B.2 Emergency Coordinator
- 2. NEI 99-01 HG7II Document No. I Rev. 0 Page 158 of 2721 ATTACHMENT IEAL BasesCategory S -System Malfunction EAL Group: Hot Conditions (NCS temperature
> 2000F); EALs inthis category are applicable only in one or more hotoperating modes.Numerous system-related equipment failure events that warrant emergency classification havebeen identified in this category.
They may pose actual or potential threats to plant safety.The events of this category pertain to the following subcategories:
- 1. Loss of Essential AC PowerLoss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may benecessary to ensure fission product barrier integrity.
This category includes loss of onsiteand offsite sources for 4160 VAC essential buses.2. Loss of Vital DC PowerLoss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may benecessary to ensure fission product barrier integrity.
This category includes loss of vitalplant 125 VDC power sources.3. Loss of Control Room Indications Certain events that degrade plant operator ability to effectively assess plant conditions within the plant warrant emergency classification.
Losses of indicators are in thissubcategory.
- 4. NCS ActivityDuring normal operation, reactor coolant fission product activity is very low. Smallconcentrations of fission products in the coolant are primarily from the fission of trampuranium in the fuel clad or minor perforations in the clad itself. Any significant increasefrom these base-line levels (2% -5% clad failures) is indicative of fuel failures and iscovered under the Fission Product Barrier Degradation category.
- However, lesser amountsof clad damage may result in coolant activity exceeding Technical Specification limits.These fission products will be circulated with the reactor coolant and can be detected bycoolant sampling.
- 5. NCS LeakageThe reactor vessel provides a volume for the coolant that covers the reactor core. Thereactor pressure vessel and associated pressure piping (reactor coolant system) togetherprovide a barrier to limit the release of radioactive material should the reactor fuel cladintegrity fail. Excessive NCS leakage greater than Technical Specification limits indicates Document No. Rev. 0 Page 159 of 272]
ATTACHMENT 1EAL Basespotential pipe cracks that may propagate to an extent threatening fuel clad, NCS andcontainment integrity.
- 6. RPS FailureThis subcategory includes events related to failure of the Reactor Protection System (RPS)to initiate and complete reactor trips. In the plant licensing basis, postulated failures of theRPS to complete a reactor trip comprise a specific set of analyzed events referred to asAnticipated Transient Without Scram (ATWS) events. For EAL classification, however,ATWS is intended to mean any trip failure event that does not achieve reactor shutdown.
IfRPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk andcould cause a threat to fuel clad, NCS and containment integrity.
- 7. Loss of Communications Certain events that degrade plant operator ability to effectively communicate with essential personnel within or external to the plant warrant emergency classification.
- 8. Containment Isolation FailureFailure of containment isolation capability (under conditions in which the containment is notcurrently challenged) warrants emergency classification.
- 9. Hazardous Event Affectinq Safety SystemsVarious natural and technological events that result in degraded plant safety systemperformance or significant visible damage warrant emergency classification under thissubcategory.
I Document No. I Rev. 0 1 Page 160 of 272]
ATTACHMENT 1EAL BasesCategory:
S -System Malfunction Subcategory:
1 -Loss of Essential AC PowerInitiating Condition:
Loss of all offsite AC power capability to essential buses for 15minutes or longerEAL:SUl.1 Unusual EventLoss of all offsite AC power capability, Table S-1, to essential 4160V buses 1(2)ETA and1 (2)ETB for > 15 min. (Note 1)Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Table S-1 AC Power SourcesOffsite:" ATC (Train A)" SATA (Train A)" ATD (Train B)" SATB (Train B)Onsite:" D/G 1(2) A (Train A)" D/G 1(2) B (Train B)Mode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
NoneBasis:MNS Basis:The 4160 VAC System provides the power requirements for operation and safe shutdown ofthe plant. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1).The essential buses are normally powered from the 6.9KV offsite power system through theirrespective 6.9KV/4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, aI Document No. Rev. 0 Page 161 of 272 ATTACHMENT 1EAL Basesstandby source of power to each 4160V essential bus is provided from the 6.9KV offsite powersystem via two separate and independent 6.9KV/4160V transformers (SATA & SATB). Thesetransformers are shared between the two units (ref. 1, 2).Each essential bus has a dedicated diesel generator (D/G 1(2) A & DIG 1(2) B) to supply anonsite emergency source of power to safe shutdown loads in the event of a loss of the normalpower source or loss of off-site power. The D/Gs will automatically start and tie onto theessential buses if the normal power source or off-site power is lost (ref. 1).An Alternate AC power source, the Standby Shutdown Diesel Generator, which providespower to the Standby Shutdown System, is located in the Standby Shutdown Facility (SSF).This AC power source must be started locally from the SSF Control Room. The SSF DieselGenerator has sufficient capability to operate equipment necessary to maintain a safeshutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event but is not credited as an AC power source byTechnical Specifications (ref. 1).The 15-minute interval was selected as a threshold to exclude transient or momentary powerlosses.NEI 99-01 Basis:This IC addresses a prolonged loss of offsite power. The loss of offsite power sourcesrenders the plant more vulnerable to a complete loss of power to AC emegeRGY-essential buses. This condition represents a potential reduction in the level of safety of the plant.For emergency classification
- purposes, "capability" means that an offsite AC power source(s) is available to the emei-eRY
-essential buses, whether or not the buses are powered from it.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofoffsite power.Escalation of the emergency classification level would be via IC SA1.MNS Basis Reference(s):
- 1. UFSAR Section 8.0 Electric Power2. AP/1(2)/A/5500/07 Loss of Electrical Power3. ECA-0.0 EP/1 (2)/A15000/ECA-0.0 Loss of All AC Power4. NEI 99-01 SUWDocument No. Rev. 0 Page 162 of 272]
ATTACHMENT 1EAL BasesCategory:
Subcategory:
S -System Malfunction 1 -Loss of Emergency AC PowerInitiating Condition:
Loss of all but one AC power source to essential buses for 15 minutesor longerEAL:SA1.1 AlertAC power capability, Table S-1, to essential 4160V buses 1(2)ETA and I(2)ETB reducedto a single power source for > 15 min. (Note 1)ANDAny additional single power source failure will result in loss of all AC power to SAFETYSYSTEMSNote 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Table S-1 AC Power SourcesOffsite:" ATC (Train A)" SATA (Train A)" ATD (Train B)" SATB (Train B)Onsite:" D/G 1(2) A (Train A)" D/G 1(2) B (Train B)Mode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional duringand following design basis events to assure:Document No. I Rev. 0 Page 163 of 272]
ATTACHMENT 1EAL Bases(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which couldresult in potential offsite exposures.
Basis:MNS Basis:For emergency classification
- purposes, "capability" means that an AC power source isavailable to the essential buses, whether or not the buses are powered from it.The 4160 VAC System provides the power requirements for operation and safe shutdown ofthe plant. The essential switchgear are buses ETA (Train A) and ETB (Train B) (ref. 1).The essential buses are normally powered from the 6.9KV offsite power system through theirrespective 6.9KV/4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, astandby source of power to each 4160V essential bus is provided from the 6.9KV offsite powersystem via two separate and independent 6.9KV/4160V transformers (SATA & SATB). Thesetransformers are shared between the two units (ref. 1, 2).Each essential bus has a dedicated diesel generator (DIG 1(2) A & D/G 1(2) B) to supply anonsite emergency source of power to safe shutdown loads in the event of a loss of the normalpower source or loss of off-site power. The D/Gs will automatically start and tie onto theessential buses if the normal power source or off-site power is lost (ref. 1).An Alternate AC power source, the Standby Shutdown Diesel Generator, which providespower to the Standby Shutdown System, is located in the Standby Shutdown Facility (SSF).This AC power source must be started locally from the SSF Control Room. The SSF DieselGenerator has sufficient capability to operate equipment necessary to maintain a safeshutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event but is not credited as an AC power source byTechnical Specifications (ref. 1).The 15-minute interval was selected as a threshold to exclude transient or momentary powerlosses. If the capability of a second source of emergency bus power is not restored within 15minutes, an Alert is declared under this EAL.NEI 99-01 Basis:This IC describes a significant degradation of offsite and onsite AC power sources such thatany additional single failure would result in a loss of all AC power to SAFETY SYSTEMS.
Inthis condition, the sole AC power source may be powering one, or more than one, train ofsafety-related equipment.
This IC provides an escalation path from IC SU1.An "AC power source" is a source recognized in AOPs and EOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.Document No. Rev. 0 Page 164 of 272]
ATTACHMENT 1EAL Bases" A loss of all offsite power with a concurrent failure of all but one emergency powersource (e.g., an onsite diesel generator).
" A loss of all offsite power and loss of all emergency power sources (e.g., onsite dieselgenerators) with a single train of emergency buses being back-fed from the unit maingenerator.
" A loss of emergency power sources (e.g., onsite diesel generators) with a single train ofemergency buses being baGk-fed from an offsite power source.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofpower.Escalation of the emergency classification level would be via IC SSI.MNS Basis Reference(s):
- 1. UFSAR Section 8.0 Electric Power2. AP/I(2)/1A5500/07 Loss of Electrical Power3. ECA-0.0 EP/I(2)/IA5000/ECA-0.0 Loss of All AC Power4. NEI 99-01 SA1IDocument No. Rev. 0 Page 165 of 272 ATTACHMENT 1EAL BasesCategory:
Subcategory:
S -System Malfunction 1 -Loss of Emergency AC PowerInitiating Condition:
Loss of all offsite power and all onsite AC power to essential buses for15 minutes or longerEAL:SSI.1 Site Area Emergency Loss of all offsite and all onsite AC power capability, Table S-1, to essential 4160V buses1(2)ETA and I(2)ETB for > 15 min. (Note 1)Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Table S-1 AC Power SourcesOffsite:" ATC (Train A)" SATA (Train A)" ATD (Train B)" SATB (Train B)Onsite:* D/G 1(2) A (Train A)" D/G 1(2) B (Train B)Mode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
NoneMNS Basis:This EAL is indicated by the loss of all offsite and onsite AC power capability (Table C-2) to4160V essential buses ETA and ETB. The essential switchgear are buses ETA (Train A) andETB (Train B) (ref. 1). For emergency classification
- purposes, "capability" means that an ACpower source is available to the essential buses, whether or not the buses are powered fromit.I Document No. I Rev. 0 Page 166 of 272]
ATTACHMENT 1EAL BasesThe essential buses are normally powered from the 6.9KV offsite power system through theirrespective 6.9KV/4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, astandby source of power to each 4160V essential bus is provided from the 6.9KV offsite powersystem via two separate and independent 6.9KV/4160V transformers (SATA & SATB). Thesetransformers are shared between the two units (ref. 1, 2).Each essential bus has a dedicated diesel generator (DIG 1(2) A & D/G 1(2) B) to supply anonsite emergency source of power to safe shutdown loads in the event of a loss of the normalpower source or loss of off-site power. The D/Gs will automatically start and tie onto theessential buses if the normal power source or off-site power is lost (ref. 1).An Alternate AC power source, the Standby Shutdown Diesel Generator, which providespower to the Standby Shutdown System, is located in the Standby Shutdown Facility (SSF).This AC power source must be started locally from the SSF Control Room. The SSF DieselGenerator has sufficient capability to operate equipment necessary to maintain a safeshutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event but is not credited as an AC power source byTechnical Specifications (ref. 1).The 15-minute interval was selected as a threshold to exclude transient or momentary powerlosses. The interval begins when both offsite and onsite AC power capability are lost.NEI 99-01 Basis:This IC addresses a total loss of AC power that compromises the performance of all SAFETYSYSTEMS requiring electric power including those necessary for emergency core cooling,containment heat removal/pressure
- control, spent fuel heat removal and the ultimate heatsink. In addition, fission product barrier monitoring capabilities may be degraded under theseconditions.
This IC represents a condition that involves actual or likely major failures of plantfunctions 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 ICs AG-1-RG1, FG1 or SG1.MNS Basis Reference(s):
- 1. UFSAR Section 8.0 Electric Power2. AP/1(2)/A/5500/07 Loss of Electrical Power3. ECA-0.0 EP/1 (2)/A/5000/ECA-0.0 Loss of All AC Power4. NEI 99-01 SS1Document No. Rev. 0 Page 167 of 272]
ATTACHMENT 1EAL BasesCategory:
Subcategory:
S -System Malfunction 1 -Loss of Essential AC PowerInitiating Condition:
Prolonged loss of all offsite and all onsite AC power to essential busesEAL:SGI.1 General Emergency Loss of all offsite and all onsite AC power capability to essential 4160V buses 1(2)ETAand 1(2)ETBAND EITHER:* Restoration of at least one essential bus in < 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is not likely (Note 1)" Core Cooling RED PATH conditions metNote 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
NoneMNS Basis:This EAL is indicated by the extended loss of all offsite and onsite AC power capability to4160V emergency buses ETA and ETB either for greater then the MNS Station Blackout(SBO) coping analysis time (4 hrs.) (ref. 1) or that has resulted in indications of an actual lossof adequate core cooling.Indication of continuing core cooling degradation is manifested by CSFST Core Cooling REDPATH conditions being met. (ref. 2).The essential buses are normally powered from the 6.9KV offsite power system through theirrespective 6.9KV/4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, astandby source of power to each 4160V essential bus is provided from the 6.9KV offsite powersystem via two separate and independent 6.9KV/4160V transformers (SATA & SATB). Thesetransformers are shared between the two units (ref. 1, 2).Each essential bus has a dedicated diesel generator (DIG 1(2) A & DIG 1(2) B) to supply anonsite emergency source of power to safe shutdown loads in the event of a loss of the normalpower source or loss of off-site power. The D/Gs will automatically start and tie onto theessential buses if the normal power source or off-site power is lost (ref. 1).Document No. Rev. 0 1 Page 168 of 272 ATTACHMENT 1EAL BasesAn Alternate AC power source, the Standby Shutdown Diesel Generator, which providespower to the Standby Shutdown System, is located in the Standby Shutdown Facility (SSF).This AC power source must be started locally from the SSF Control Room. The SSF DieselGenerator has sufficient capability to operate equipment necessary to maintain a safeshutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event (ref. 3).Four hours is the station blackout coping time (ref 2).Indication of continuing core cooling degradation must be based on fission product barriermonitoring with particular emphasis on Emergency Coordinator judgment as it relates toimminent Loss or Potential Loss of fission product barriers and degraded ability to monitorfission product barriers.
Indication of continuing core cooling degradation is manifested byCSFST Core Cooling RED PATH conditions being met (ref. 2). Specifically, Core Cooling REDPATH conditions exist if either core exit T/Cs are reading greater than or equal to 1200OF orsubcooling is 0°F AND no NC pumps are on AND core exit T/Cs are reading greater than orequal to 700°F AND Reactor Vessel Lower Range level less than or equal to 39% (ref. 2).NEI 99-01 Basis:This IC addresses a prolonged loss of all power sources to AC emengeR*y-essential buses. Aloss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electricpower including those necessary for emergency core cooling, containment heatremoval/pressure
- control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers.
In addition, fission product barrier monitoring capabilities may be degraded under these conditions.
The EAL should require declaration of a General Emergency prior to meeting the thresholds for IC FG1. This will allow additional time for implementation of offsite protective actions.Escalation of the emergency classification from Site Area Emergency will occur if it isprojected that power cannot be restored to at least one AC emeFgenY-essential bus by theend of the analyzed station blackout coping period. Beyond this time, plant responses andevent trajectory are subject to greater uncertainty, and there is an increased likelihood ofchallenges to multiple fission product barriers.
The estimate for restoring at least one emeFgeny-essential bus should be based on a realistic appraisal of the situation.
Mitigation actions with a low probability of success should not beused as a basis for delaying a classification upgrade.
The goal is to maximize the timeavailable to prepare for, and implement, protective actions for the public.The EAL will also require a General Emergency declaration if the loss of AC power results inparameters that indicate an inability to adequately remove decay heat from the core.MNS Basis Reference(s):
- 1. UFSAR Section 8.4.2 Station Blackout Duration2. EP/1(2)/A/5000/F-0 Critical Safety Function Status Tress -Core Cooling3. UFSAR Section 8.0 Electric Power4. AP/1(2)/A/5500/07 Loss of Electrical PowerDocument No. Rev. 0 Page 169 of 272 ATTACHMENT 1EAL Bases5. ECA-0.0 EP/I1(2)/A/5000/ECA-0.0 Loss of All AC Power6. NEI 99-01 SG1I Document No. I Rev. 0 1 Page 170o,272 ATTACHMENT 1EAL BasesCategory:
S -System Malfunction Subcategory:
1 -Loss of Essential AC PowerInitiating Condition:
Loss of all essential AC and vital DC power sources for 15 minutes orlongerEAL:SG1.2 General Emergency Loss of all offsite and all onsite AC power capability, Table S-1, to essential 4160V buses1(2)ETA and 1(2)ETB for >- 15 min.ANDLoss of all 125 VDC power based on battery bus voltage indications
< 105 VDC on bothvital DC buses EVDA and EVDD for -> 15 min.(Note 1)Note 1:The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Table S-1 AC Power SourcesOffsite:" ATC (Train A)" SATA (Train A)" ATD (Train B)* SATB (Train B)Onsite:" D/G 1(2) A (Train A)" D/G 1(2) B (Train B)Mode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDocument No. Rev. 0 Page 171 of 272:]
ATTACHMENT 1EAL BasesDefinition(s):
NoneMNS Basis:This EAL is indicated by the loss of all offsite and onsite emergency AC power capability to4160V emergency buses ETA and ETB for greater than 15 minutes in combination withdegraded vital DC power voltage.
This EAL addresses operating experience from the March2011 accident at Fukushima Daiichi.The essential buses are normally powered from the 6.9KV offsite power system through theirrespective 6.9KV/4160V Station Auxiliary Transformers (1ATC & 1ATD). Additionally, astandby source of power to each 4160V essential bus is provided from the 6.9KV offsite powersystem via two separate and independent 6.9KV/4160V transformers (SATA & SATB). Thesetransformers are shared between the two units (ref. 1, 2).Each essential bus has a dedicated diesel generator (D/G 1(2) A & D/G 1(2) B) to supply anonsite emergency source of power to safe shutdown loads in the event of a loss of the normalpower source or loss of off-site power. The D/Gs will automatically start and tie onto theessential buses if the normal power source or off-site power is lost (ref. 1).An Alternate AC power source, the Standby Shutdown Diesel Generator, which providespower to the Standby Shutdown System, is located in the Standby Shutdown Facility (SSF).This AC power source must be started locally from the SSF Control Room. The SSF DieselGenerator has sufficient capability to operate equipment necessary to maintain a safeshutdown condition for the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO event (ref. 1).The 125 VDC electrical power system consists of two independent and redundant safetyrelated Class 1 E DC electrical power subsystems (Train A or EVDA, and Train B or EVDD).Each subsystem consists of two channels of 125 VDC batteries (each battery 100% capacity),
the associated battery charger(s) for each battery, and all the associated control equipment and interconnecting cabling.
(ref. 1).The Train A and Train B DC electrical power subsystems provide the control power for its associated Class 1E AC power load group, 4.16 kV switchgear, and 600 V load centers.
The DC electrical powersubsystems also provide DC electrical power to the inverters, which in turn power the AC vital buses.(ref. 1, 3).The minimum battery discharge voltage (requiring opening the degraded battery outputbreaker) is 105 VDC (ref. 1, 3).NEI-9901 Basis:This IC addresses a concurrent and prolonged loss of both essential AC and Vital DC power.A loss of all essential AC power compromises the performance of all SAFETY SYSTEMSrequiring 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 ofDocument No. I Rev. 0 1 Page 172 of 272 1 ATTACHMENT 1EAL Basesvital DC power compromises the ability to monitor and control SAFETY SYSTEMS.
Asustained loss of both essential AC and vital DC power will lead to multiple challenges tofission 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 aremet.MNS Basis Reference(s):
- 1. UFSAR Section 8.0 Electric Power2. AP/1 (2)/A/5500/07 Loss of Electrical Power3 AP/1 (2)/A/5500/15 Loss of Vital or Aux Control Power4. ECA-0.0 EP/1 (2)/A/5000/ECA-0.0 Loss of All AC Power5. NEI 99-01 SG8I Document No. I Rev. 0 Page 173 of 272 1 ATTACHMENT 1EAL BasesCategory:
Subcategory:
S -System Malfunction 2 -Loss of Vital DC PowerInitiating Condition:
Loss of all vital DC power for 15 minutes or longerEAL:SS2.1 Site Area Emergency Loss of all 125 VDC power based on battery bus voltage indications
< 105 VDC on bothvital DC buses EVDA and EVDD for > 15 min (Note 1)Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
NoneMNS Basis:The 125 VDC electrical power system consists of two independent and redundant safetyrelated Class 1 E DC electrical power subsystems (Train A or EVDA, and Train B or EVDD).Each subsystem consists of two channels of 125 VDC batteries (each battery 100% capacity),
the associated battery charger(s) for each battery, and all the associated control equipment and interconnecting cabling.
(ref. 1).The Train A and Train B DC electrical power subsystems provide the control power for its associated Class 1E AC power load group, 4.16 kV switchgear, and 600 V load centers.
The DC electrical powersubsystems also provide DC electrical power to the inverters, which in turn power the AC vital buses.(ref. 1,2).The minimum battery discharge voltage (requiring opening the degraded battery outputbreaker) is 105 VDC (ref. 1, 2).NEI 99-01 Basis:This IC addresses a loss of vital DC power which compromises the ability to monitor andcontrol SAFETY SYSTEMS.
In modes above Cold Shutdown, this condition involves a majorfailure 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 ICs AG-1-RG1, FG1 or SG8SG1.Document No. Rev. 0 Page 174 of 272 ATTACHMENT 1EAL BasesMNS Basis Reference(s):
1.23.UFSAR Section 8.0 Electric PowerAP/1 (2)/A/5500/15 Loss of Vital or Aux Control PowerNEI 99-01 SS8I Document No. I Rev. 0 F Page 175 of 272]
ATTACHMENT 1EAL BasesCategory:
Subcategory:
S -System Malfunction 3 -Loss of Control Room Indications Initiating Condition:
UNPLANNED loss of Control Room indications for 15 minutes orlongerEAL:SU3.1 Unusual EventAn UNPLANNED event results in the inability to monitor one or more Table S-2parameters from within the Control Room for > 15 min. (Note 1)Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Table S-2 Safety System Parameters
" Level in at least one S/G" Auxiliary feed flow in at least one S/GMode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
UNPLANNED
-A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient.
The cause of the parameter changeor event may be known or unknown.MNS Basis:SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room througha combination of hard control panel indicators as well as computer based information systems.The Operator Aid Computer (OAC), which displays SPDS required information, serves as aredundant compensatory indicator which may be utilized in lieu of normal Control Roomindicators (ref. 1, 2).I Document No. I Rev. 0 Page 176 of 272 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses the difficulty associated with monitoring normal plant conditions without theability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition isa precursor to a more significant event and represents a potential degradation in the level ofsafety of the plant.As used in this EAL, an "inability to monitor" means that values for one or more of the listedparameters cannot be determined from within the Control Room. This situation would requirea loss of all of the Control Room sources for the given parameter(s).
For example, the reactorpower level cannot be determined from any analog, digital and recorder source within theControl Room.An event involving a loss of plant indications, annunciators and/or display systems isevaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) todetermine 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 safetyfunctions of reactivity
- control, core cooling [PWq / RPV love.,ol
[B'RI and NCS heat removal.The loss of the ability to determine one or more of these parameters from within the ControlRoom is considered to be more significant than simply a reportable condition.
In addition, if allindication sources for one or more of the listed parameters are lost, then the ability todetermine the values of other SAFETY SYSTEM parameters may be impacted as well. Forexample, if the value for reactor vessel level [PWR]! RPV wat.. level [BWR. cannot bedetermined from the indications and recorders on a main control board, the SPDS or the plantcomputer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofindication.
Escalation of the emergency classification level would be via IC SA2SA3.MNS Basis Reference(s):
- 1. UFSAR Section 7.5 Safety-Related Display Instrumentation
- 2. OP/1 (2)/A/6100/SD-2 Cooldown to 400 Degrees F3. NEI 99-01 SU2Document No. Rev. 0 Page 177 of 2721 ATTACHMENT 1EAL BasesCategory:
Subcategory:
S -System Malfunction 3 -Loss of Control Room Indications Initiating Condition:
UNPLANNED loss of Control Room indications for 15 minutes orlonger with a significant transient in progressEAL:SA3.1 AlertAn UNPLANNED event results in the inability to monitor one or more Table S-2parameters from within the Control Room for > 15 min. (Note 1)ANDAny significant transient is in progress, Table S-3Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Table S-2 Safety System Parameters
- Level in at least one S/G* Auxiliary or emergency feed flowTable S-3 Significant Transients
" Reactor trip" Runback > 25% thermal power" Electrical load rejection
> 25%electrical load" Safety injection actuation Mode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownI Document No. Rev. 0 Page 178 of 272 ATTACHMENT 1EAL BasesDefinition(s):
UNPLANNED
-A parameter change or an event that is not 1) the result of an intendedevolution or 2) an expected plant response to a transient.
The cause of the parameter changeor event may be known or unknown.MNS Basis:SAFETY SYSTEM parameters listed in Table S-1 are monitored in the Control Room througha combination of hard control panel indicators as well as computer based information systems.The Operator Aid Computer (OAC), which displays SPDS required information, serves as aredundant compensatory indicator which may be utilized in lieu of normal Control Roomindicators (ref. 1,2).Significant transients are listed in Table S-2 and include response to automatic or manuallyinitiated functions such as reactor trips, runbacks involving greater than 25% thermal powerchange, electrical load rejections of greater than 25% full electrical load or SI injection actuations.
NEI 99-01 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 theControl 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 theplant.As used in this EAL, an "inability to monitor" means that values for one or more of the listedparameters cannot be determined from within the Control Room. This situation would requirea loss of all of the Control Room sources for the given parameter(s).
For example, the reactorpower level cannot be determined from any analog, digital and recorder source within theControl Room.An event involving a loss of plant indications, annunciators and/or display systems isevaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1 022) todetermine 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 safetyfunctions of reactivity
- control, core cooling [PWR] I RPV level [BW]4 and NCS heat removal.The loss of the ability to determine one or more of these parameters from within the ControlRoom is considered to be more significant than simply a reportable condition.
In addition, if allindication sources for one or more of the listed parameters are lost, then the ability todetermine the values of other SAFETY SYSTEM parameters may be impacted as well. Forexample, if the value for reactor vessel level [P.R ! RPV water l.vol [B-AW cannot beDocument No. I Rev. 0 Page 179 of 272 ATTACHMENT 1EAL Basesdetermined from the indications and recorders on a main control board, the SPDS or the plantcomputer, the availability of other parameter values may be compromised as well.Fifteen minutes was selected as a threshold to exclude transient or momentary losses ofindication.
Escalation of the emergency classification level would be via ICs FS1 or IC AS4RS1MNS Basis Reference(s):
- 1. UFSAR Section 7.5 Safety-Related Display Instrumentation
- 2. OP/1(2)/A/6100/SD-2 Cooldown to 400 Degrees F3. NEI 99-01 SA2I Document No. Rev. 0 Page 180 of 272 ATTACHMENT 1EAL BasesCategory:
S -System Malfunction Subcategory:
4 -NCS ActivityInitiating Condition:
NCS activity greater than Technical Specification allowable limitsEAL:SU4.1 Unusual EventDose Equivalent 1-131 activity
> 1 [tCi/gmORDose Equivalent Xe-133 activity
> 280 tCi/gmMode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
NoneMNS Basis:The specific iodine activity is limited to -1.0 pCi/gm Dose Equivalent 1-131. The specific Xe-133 activity is limited to -280 pCi/gm Dose Equivalent XE-133 (ref 1, 2).NEI 99-01 Basis:This IC addresses a reactor coolant activity value that exceeds an allowable limit specified inTechnical Specifications.
This condition is a precursor to a more significant event andrepresents a potential degradation of the level of safety of the plant.Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category A-R ICs.MNS Basis Reference(s):
- 1. MNS Technical Specifications section 3.4.16 RCS Specific Activity2. MNS Technical Specifications section 3.4.16 RCS Specific Activity Bases3. NEI 99-01 SU3Document No. Rev. 0 Page 181 of 272:
ATTACHMENT 1EAL BasesCategory:
Subcategory:
S -System Malfunction 5 -NCS LeakageInitiating Condition:
NCS leakage for 15 minutes or longerEAL:SU5.1 Unusual EventNCS unidentified or pressure boundary leakage > 10 gpm for -15 min.ORNCS identified leakage > 25 gpm for > 15 min.ORLeakage from the NCS to a location outside containment
> 25 gpm for > 15 min.(Note 1)Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
NoneMNS Basis:Identified leakage includes leakage such as that from pump seals or valve packing (exceptreactor coolant pump (NCP) seal water injection or leakoff),
that is captured and conducted tocollection systems or a sump or collecting tank, leakage into the containment atmosphere fromsources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary leakage; or NCS leakagethrough a steam generator to the secondary system (primary to secondary leakage)
(ref. 1).Unidentified leakage is all leakage (except NCP seal water injection or leakoff) that is notidentified leakage (ref. 1).Pressure Boundary leakage is leakage (except primary to secondary leakage) through anonisolable fault in an NCS component body, pipe wall, or vessel wall (ref. 1)NCS leakage outside of the containment that is not considered identified or unidentified leakage per Technical Specifications includes leakage via interfacing systems such as NCS tothe Component Cooling Water (KC), or systems that directly see NCS pressure outsidecontainment such as Chemical
& Volume Control System (NV), Nuclear Sampling system(NM) and Residual Heat Removal (ND) system (when in the shutdown cooling mode).I Document No. Rev. 0 Page 182 of 272 ATTACHMENT 1EAL BasesEscalation of this EAL to the Alert level is via Category F, Fission Product Barrier Degradation, EAL FAI.1.NEI 99-01 Basis:This IC addresses RGS-NCS leakage which may be a precursor to a more significant event.In this case, RGSNCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to bea potential degradation of the level of safety of the plant.EAL #1 and EAL -2The first and second EAL conditions are focused on a loss of mass fromthe NCS due to "unidentified leakage",
"pressure boundary leakage" or "identified leakage" (asthese leakage types are defined in the plant Technical Specifications).
EAL-#3The thirdcondition addresses an NCS mass loss caused by an UNISOLABLE leak through aninterfacing system. These EAL-s-conditions thus apply to leakage into the containment, asecondary-side system (e.g., steam generator tube leakage ii;a PWR) or a location outside ofcontainment.
The leak rate values for each EAL-condition were selected because they are usuallyobservable with normal Control Room indications.
Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation).
EAL- #! The firstcondition uses a lower value that reflects the greater significance of unidentified or pressureboundary leakage.The release of mass from the RGS-NCS due to the as-designed/expected operation of a reliefvalve does not warrant an emergency classification.
For PWRs--aAn emergency classification would be required if a mass loss is caused by a relief valve that is not functioning asdesigned/expected (e.g., a relief valve sticks open and the line flow cannot be isolated).
-BWRs,-a-rtuck opon Safety Reieof Valve (SRV) or SRV leakage is not Gensidered efitheridentified Or unidentified leakage by Tochnical Specificatiens and, therefore, is not applicable
÷,.- t1,,;s The 15-minute threshold duration allows sufficient time for prompt operator actions to isolatethe leakage, if possible.
Escalation of the emergency classification level would be via ICs of Recognition Category A-Ror F.MNS Basis Reference(s):
- 1. MNS Technical Specifications Definitions section 1.12. NEI 99-01 SU4I Document No. Rev. 0 Page 183 of 272 1 ATTACHMENT 1EAL BasesCategory:
Subcategory:
S -System Malfunction 6 -RPS FailureInitiating Condition:
Automatic or manual trip fails to shut down the reactorEAL:SU6.1 Unusual EventAn automatic trip did not shut down the reactor as indicated by reactor power > 5% afterany RPS setpoint is exceededANDA subsequent automatic trip or manual trip action taken at the reactor control console(manual reactor trip switches or turbine manual trip) is successful in shutting down thereactor as indicated by reactor power < 5% (Note 8)Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidlyinserted into the core, and does not include manually driving in control rods or implementation of boroninjection strategies.
Mode Applicability:
1 -Power Operations Definition(s):
NoneMNS Basis:The first condition of this EAL identifies the need to cease critical reactor operations byactuation of the automatic Reactor Protection System (RPS) trip function.
A reactor trip isautomatically initiated by the RPS when certain continuously monitored parameters exceedpredetermined setpoints (ref. 1).Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear powerpromptly drops to a fraction of the original power level and then decays to a level severaldecades less with a negative startup rate. The reactor power drop continues until reactorpower reaches the point at which the influence of source neutrons on reactor power starts tobe observable.
A predictable post-trip response from an automatic reactor trip signal shouldtherefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful trip has therefore occurred whenthere is sufficient rod insertion from the trip of RPS to bring the reactor power below theimmediate shutdown decay heat level of 5% (ref. 2, 3, 4).For the purposes of emergency classification, successful manual trip actions are those whichcan be quickly performed from the reactor control console (i.e., manual trip switches or turbineFDocument No. 7 Rev. 0 1 Page 184 of 272 ATTACHMENT 1EAL Basestrip). Reactor shutdown achieved by use of other trip actions specified in EPII(2)/AI5000IFR-S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do notconstitute a successful manual trip (ref. 4).Following any automatic RPS trip signal, EP/1 (2)/A/5000/E-0 (ref. 2) and EP/1 (2)/A/5000/FR-S.1 (ref. 3) prescribe insertion of redundant manual trip signals to back up the automatic RPStrip function and ensure reactor shutdown is achieved.
Even if the first subsequent manual tripsignal inserts all control rods to the full-in position immediately after the initial failure of theautomatic trip, the lowest level of classification that must be declared is an Unusual Event (ref.4).In the event that the operator identifies a reactor trip is imminent and initiates a successful manual reactor trip before the automatic RPS trip setpoint is reached, no declaration isrequired.
The successful manual trip of the reactor before it reaches its automatic trip setpointor reactor trip signals caused by instrumentation channel failures do not lead to a potential fission product barrier loss. However, if subsequent manual reactor trip actions fail to reducereactor power below 5%, the event escalates to the Alert under EAL SA6.1.If by procedure, operator actions include the initiation of an immediate manual trip following receipt of an automatic trip signal and there are no clear indications that the automatic tripfailed (such as a time delay following indications that a trip setpoint was exceeded),
it may bedifficult to determine if the reactor was shut down because of automatic trip or manual actions.If a subsequent review of the trip actuation indications reveals that the automatic trip did notcause the reactor to be shut down, then consideration should be given to evaluating the fuelfor potential damage, and the reporting requirements of 50.72 should be considered for thetransient event.NEI 99-01 Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor(trip [PW.4R] / Gcram [BWRI) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [{P-WRJSGram-FnK44]/R is successful in shutting down the reactor.
This event is a precursor to a moresignificant condition and thus represents a potential degradation of the level of safety of theplant.Following the failure on an automatic reactor (trip [PW.] i Sc...rA [&--R]),
operators willpromptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g.,initiate a manual reactor (trip [PWR] i scram. [B,--R])).
If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within thecapabilities of the plant's decay heat removal systems.If an initial manual reactor (trip [PW,] i scram is unsuccessful, operators will promptlytake manual action at another location(s) on the reactor control consoles to shutdown thereactor (e.g., initiate a manual reactor (trip ,P4r'9, ]-/-,a..
[using a different switch).Depending upon several factors, the initial or subsequent effort to manually (trip [P-WR4..FamfBfiW) the reactor, or a concurrent plant condition, may lead to the generation of anDocument No. Rev. 0 Page 185 of 272 ATTACHMENT 1EAL Basesautomatic reactor (trip [PWI / signal. If a subsequent manual or automatic (trip [PWRI / scram. [B.4,..)
is successful in shutting down the reactor, core heat generation willquickly fall to a level within the capabilities of the plant's decay heat removal systems.A manual action at the reactor control consoles is any operator action, or set of actions, whichcauses the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scF, )). This action does not include manually driving in control rods orimplementation of boron injection strategies.
Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be"at the reactor control consoles".T-aki.g tho Modo S,,wit.h to SHUTDOW.A^,N i6considorod to bo a mnanual crGamn actien. [B3WRIThe plant response to the failure of an automatic or manual reactor (trip [PWR]-/sG
,BWR.) will vary based upon several factors including the reactor power level prior to theevent, availability of the condenser, performance of mitigation equipment and actions, otherconcurrent plant conditions, etc. If subsequent operator manual actions taken at the reactorcontrol consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA5SA6. Depending upon the plantresponse, escalation is also possible via IC FA1. Absent the plant conditions needed to meeteither IC SA&SA6 or FA1, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor (trip [PWR, / scram signal be generated as a result of plant work(e.g., RPS setpoint testing),
the following classification guidance should be applied." If the signal causes a plant transient that should have included an automatic reactor(trip [PL.r-R]
/ cr6m .rr ['RJ-) and the RPS fails to automatically shutdown the reactor,then this IC and the EALs are applicable, and should be evaluated.
" If the signal does not cause a plant transient and the (trip[,r,
/ scr m ... [B,'R]) failureis determined through other means (e.g., assessment of test results),
then this IC andthe EALs are not applicable and no classification is warranted.
MNS Basis Reference(s):
- 1. MNS Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation
- 2. EP/1 (2)/IA5000/E-0 Reactor Trip or Safety Injection
- 3. EP/1 (2)/A/5000/F-0 Critical Safety Function Status Trees -Subcriticality
- 4. EP/1 (2)/IA5000/FR-S.1 Response to Nuclear Power Generation/ATWS
- 5. NEI 99-01 SU5I Document No. I Rev. 0 Page 186 of 272 1 ATTACHMENT 1EAL BasesCategory:
Subcategory:
S -System Malfunction 6 -RPS FailureInitiating Condition:
Automatic or manual trip fails to shut down the reactorEAL:SU6.2 Unusual EventA manual trip did not shut down the reactor as indicated by reactor power > 5% after anymanual trip action was initiated ANDA subsequent automatic trip or manual trip action taken at the reactor control console(manual reactor trip switches or turbine manual trip) is successful in shutting down thereactor as indicated by reactor power < 5% (Note 8)Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidlyinserted into the core, and does not include manually driving in control rods or implementation of boroninjection strategies.
Mode Applicability:
1 -Power Operations Definition(s):
NoneMNS Basis:This EAL addresses a failure of a manually initiated trip in the absence of having exceeded anautomatic RPS trip setpoint and a subsequent automatic or manual trip is successful inshutting down the reactor (reactor power < 5%). (ref. 1).Following a successful reactor trip, rapid insertion of the control rods occurs. Nuclear powerpromptly drops to a fraction of the original power level and then decays to a level severaldecades less with a negative startup rate. The reactor power drop continues until reactorpower reaches the point at which the influence of source neutrons on reactor power starts tobe observable.
A predictable post-trip response from a manual reactor trip signal shouldtherefore consist of a prompt drop in reactor power as sensed by the nuclear instrumentation and a lowering of power into the source range. A successful trip has therefore occurred whenthere is sufficient rod insertion from the trip of RPS to bring the reactor power below theimmediate shutdown decay heat level of 5% (ref. 2, 3 4).For the purposes of emergency classification, successful manual trip actions are those whichcan be quickly performed from the reactor control console (i.e., manual trip switches or turbinetrip). Reactor shutdown achieved by use of other trip actions specified in EP/1 (2)/A/5000/FR-FDocument No. I Rev. 0 1 Page 187 of 272 1 ATTACHMENT 1EAL BasesS.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do notconstitute a successful manual trip (ref. 4).If both subsequent automatic and subsequent manual reactor trip actions in the Control Roomfail to reduce reactor power below the power associated with the safety system design (< 5%)following a failure of an initial manual trip, the event escalates to an Alert under EAL SA6.1NEI 99-01 Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor(trip [PWRI1 scram [914Mr) that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic (trip [LPRI-Ga.Am r [W4R]R is successful in shutting down the reactor.
This event is a precursor to a moresignificant condition and thus represents a potential degradation of the level of safety of theplant.Following the failure on an automatic reactor (trip [PWI4R] / cr.am, [.WRJ), operators willpromptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g.,initiate a manual reactor (trip [PWR] / crm-,, [...,. If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within thecapabilities of the plant's decay heat removal systems.If an initial manual reactor (trip,[rDA]
/ ,,cr]am [BWR) is unsuccessful, operators will promptlytake manual action at another location(s) on the reactor control consoles to shutdown thereactor (e.g., initiate a manual reactor (trip / scram [BW.])using a different switch).Depending upon several factors, the initial or subsequent effort to manually (tip-fPR4 sGrFam [-BW4R) the reactor, or a concurrent plant condition, may lead to the generation of anautomatic reactor (trip [PW].R] / .cram [BKr4,) signal. If a subsequent manual or automatic (trip [PI4.R] i scram [9144R])
is successful in shutting down the reactor, core heat generation willquickly fall to a level within the capabilities of the plant's decay heat removal systems.A manual action at the reactor control consoles is any operator action, or set of actions, whichcauses the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor(trip[P'R]
/ crGamn [814rM)).
This action does not include manually driving in control rods orimplementation of boron injection strategies.
Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be"at the reactor control consoles".
Taking the RoactOr Mo-do SWitc9h to_ SHUTDOWN is, considoroAd-to9 be a m~anual Scramn acion.The plant response to the failure of an automatic or manual reactor (trip LPWRD 1 SG,/"[1WRI) will vary based upon several factors including the reactor power level prior to theevent, availability of the condenser, performance of mitigation equipment and actions, otherconcurrent plant conditions, etc. If subsequent operator manual actions taken at the reactorcontrol consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC SA6SA6. Depending upon the plantDocument No. Rev. 0 Page 188 of 272 ATTACHMENT 1EAL Basesresponse, escalation is also possible via IC FA1. Absent the plant conditions needed to meeteither IC SA6-SA6 or FAI, an Unusual Event declaration is appropriate for this event.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Should a reactor (trip [PR] cram. [/. WR]) signal 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(trip[,r,
/ .cr , .'am [,--,,R])
and the RPSRTS fails to automatically shutdown thereactor, then this IC and the EALs are applicable, and should be evaluated.
o If the signal does not cause a plant transient and the (trip [PW.rDi.
cr..am [....-R])
failureis determined through other means (e.g., assessment of test results),
then this IC andthe EALs are not applicable and no classification is warranted.
MNS Basis Reference(s):
- 1. MNS Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation
- 2. EP/I(2)/A/5000/E-0 Reactor Trip or Safety Injection
- 3. EP/I(2)/A/5000/F-0 Critical Safety Function Status Trees -Subcriticality
- 4. EP/I(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS
- 5. NEI 99-01 SU5I Document No. Rev. 0 Page 189 of 272 1 ATTACHMENT 1EAL BasesCategory:
Subcategory:
Initiating Condition:
S -System Malfunction 2 -RPS FailureAutomatic or manual trip fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are notsuccessful in shutting down the reactorEAL:SA6.1 AlertAn automatic or manual trip fails to shut down the reactor as indicated by reactor power> 5%ANDManual trip actions taken at the reactor control console (manual reactor trip switches orturbine manual trip) are not successful in shutting down the reactor as indicated by reactorpower 2 5% (Note 8)Note 8: A manual trip action is any operator action, or set of actions, which causes the control rods to be rapidlyinserted into the core, and does not include manually driving in control rods or implementation of boroninjection strategies.
Mode Applicability:
1 -Power Operations Definition(s):
NoneMNS Basis:This EAL addresses any automatic or manual reactor trip signal that fails to shut down thereactor followed by a subsequent manual trip that fails to shut down the reactor to an extentthe reactor is producing energy in excess of the heat load for which the safety systems weredesigned.
For the purposes of emergency classification, successful manual trip actions are those whichcan be quickly performed from the reactor control console (i.e., manual trip switches or turbinetrip). Reactor shutdown achieved by use of other trip actions specified in EP/I (2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS (such as depressing manual pushbutton on turbine control panel, emergency boration or manually driving control rods) do notconstitute a successful manual trip (ref. 4).5% rated power is a minimum reading on the power range scale that indicates continued power production.
It also approximates the decay heat which the shutdown systems weredesigned to remove and is indicative of a condition requiring immediate response to preventFDocument No. I Rev. 0 1 Page 190 of 272 ATTACHMENT 1EAL Basessubsequent core damage. Below 5%, plant response will be similar to that observed during anormal shutdown.
Nuclear instrumentation can be used to determine if reactor power isgreater than 5 % power (ref. 1).Escalation of this event to a Site Area Emergency would be under EAL SS6.1 or Emergency Coordinator judgment.
NEI 99-01 Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor(trip [PWR]I / cr.am ,.r). that results in a reactor shutdown, and subsequent operatormanual actions taken at the reactor control consoles to shutdown the reactor are alsounsuccessful.
This condition represents an actual or potential substantial degradation of thelevel of safety of the plant. An emergency declaration is required even if the reactor issubsequently shutdown by an action taken away from the reactor control consoles since thisevent entails a significant failure of the RPS.A manual action at the reactor control console is any operator action, or set of actions, whichcauses the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor(trip[rA4R]
/ cr.am [BWR])).
This action does not include manually driving in control rods orimplementation of boron injection strategies.
If this action(s) is unsuccessful, operators wouldimmediately pursue additional manual actions at locations away from the reactor controlconsoles (e.g., locally opening breakers).
Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be"at the reactor control consoles".Taking the ReactGo Modo SWitch to isconsidorod to be a manual crGamn action. [9WR]The plant response to the failure of an automatic or manual reactor (trip fPWR] i 6G--WR]J) will vary based upon several factors including the reactor power level prior to theevent, availability of the condenser, performance of mitigation equipment and actions, otherconcurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough tocause a challenge to the core cooling [PWR- / RPV wat.. le'vel [,. WR or , GS NCS heatremoval safety functions, the emergency classification level will escalate to a Site AreaEmergency via IC SS66. Depending upon plant responses and symptoms, escalation is alsopossible via IC FSI. Absent the plant conditions needed to meet either IC SS66 or FS1, anAlert declaration is appropriate for this event.It is recognized that plant responses or symptoms may also require an Alert declaration inaccordance with the Recognition Category F ICs; however, this IC and EAL are included toensure a timely emergency declaration.
A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
MNS Basis Reference(s):
- 1. MNS Technical Specifications section 3.3.1 Reactor Trip System (RTS) Instrumentation
- 2. EP/1(2)/A/5000/E-0 Reactor Trip or Safety Injection Document No. I Rev. 0 Page 191 of 272 ATTACHMENT 1EAL Bases3. EP/1(2)/A/5000/F-0 Critical Safety Function Status Trees -Subcriticality
- 4. EP/1 (2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS
- 5. NEI 99-01 SA5I Document No. Rev. 0 Page 192 of 272]
ATTACHMENT 1EAL BasesCategory:
Subcategory:
Initiating Condition:
EAL:S -System Malfunction 2 -RPS FailureInability to shut down the reactor causing a challenge to core coolingor NCS heat removalSS6.1 Site Area Emergency An automatic or manual trip fails to shut down the reactor as indicated by reactor power-5%ANDAll actions to shut down the reactor are not successful as indicated by reactor power> 5%AND EITHER:* Core Cooling RED PATH conditions met* Heat Sink RED PATH conditions metMode Applicability:
1 -Power Operations Definition(s):
NoneMNS Basis:This EAL addresses the following:
" Any automatic reactor trip signal followed by a manual trip that fails to shut down thereactor to an extent the reactor is producing energy in excess of the heat load for whichthe safety systems were designed (EAL SA6.1), and* Indications that either core cooling is extremely challenged or heat removal is extremely challenged.
The combination of failure of both front line and backup protection systems to function inresponse to a plant transient, along with the continued production of heat, poses a directthreat to the Fuel Clad and NCS barriers.
Reactor shutdown achieved by use of EP/I(2)/A/5000/FR-S.1 Response to Nuclear PowerGeneration/ATWS (such as depressing manual pushbutton on turbine control panel,emergency boration or manually driving control rods) are also credited as a successful manualDocument No. Rev. 0 Page 193 of 272 ATTACHMENT 1EAL Basestrip provided reactor power can be reduced below 5% before indications of an extremechallenge to either core cooling or heat removal exist (ref. 1, 4).5% rated power is a minimum reading on the power range scale that indicates continued power production.
It also approximates the decay heat which the shutdown systems weredesigned to remove and is indicative of a condition requiring immediate response to preventsubsequent core damage. Below 5%, plant response will be similar to that observed during anormal shutdown.
Nuclear instrumentation can be used to determine if reactor power isgreater than 5% power (ref. 1, 4).Indication of continuing core cooling degradation is manifested by CSFST Core Cooling REDPATH conditions being met (ref. 2). Specifically, Core Cooling RED PATH conditions exist ifeither core exit T/Cs are reading greater than or equal to 1200OF or subcooling is 0°F AND noNC pumps are on AND core exit T/Cs are reading greater than or equal to 700°F AND ReactorVessel Lower Range level less than or equal to 39% (ref. 2).Indication of inability to adequately remove heat from the NCS is manifested by CSFST HeatSink RED PATH conditions being met (ref. 2). Specifically, Heat Sink RED PATH conditions exist if narrow range level in at least on steam generator is not greater than or equal to 11 %(32% ACC) and total feedwater flow to the intact steam generators is less than or equal to 450gpm. (ref. 3).NEI 99-01 Basis:This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor(trip [PW,] / scram r [BW) that results in a reactor shutdown, all subsequent operator actionsto manually shutdown the reactor are unsuccessful, and continued power generation ischallenging the capability to adequately remove heat from the core and/or the RG&NCS. Thiscondition will lead to fuel damage if additional mitigation actions are unsuccessful and thuswarrants the declaration of a Site Area Emergency.
In some instances, the emergency classification resulting from this IC/EAL may be higher thanthat resulting from an assessment of the plant responses and symptoms against theRecognition Category F ICs/EALs.
This is appropriate in that the Recognition Category FICs/EALs do not address the additional threat posed by a failure to shut down the reactor.The inclusion of this IC and EAL ensures the timely declaration of a Site Area Emergency inresponse to prolonged failure to shutdown the reactor.A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.
Escalation of the emergency classification level would be via IC AGI-RG1 or FGI.MNS Basis Reference(s):
- 1. EP/1(2)/N5000/F-0 Critical Safety Function Status Trees -Subcriticality
- 2. EP/1(2)/A15000/F-0 Critical Safety Function Status Tress -Core Cooling3. EP/1(2)/A/5000/F-0 Critical Safety Function Status Tress -Heat SinkDocument No. I Rev. 0 1 Page 194 of 272 ATTACHMENT 1EAL Bases4. EP/1(2)/A/5000/FR-S.1 Response to Nuclear Power Generation/ATWS
- 5. NEI 99-01 SS5I Document No. I Rev. 0 1 Page 195 of 272]
ATTACHMENT 1EAL BasesCategory:
S -System Malfunction Subcategory:
7 -Loss of Communications Initiating Condition:
Loss of all onsite or offsite communications capabilities EAL:SU7.1 Unusual EventLoss of all Table S-4 onsite communication methodsORLoss of all Table S-4 ORO communication methodsORLoss of all Table S-4 NRC communication methodsTable S-4 Communication MethodsSystem Onsite ORO NRCPublic Address XInternal Telephones XOnsite Radios XDEMNET XOffsite Radio System XCommercial Telephones X XNRC Emergency Telephone System (ETS) XMode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
NoneDocument No. Rev. 0 Page 196 of 272]
ATTACHMENT 1EAL BasesMNS Basis:Onsite/offsite communications include one or more of the systems listed in Table S-4 (ref. 1).Public Address SystemThe McGuire Nuclear Station public address system provides paging and party linecommunications between stations located throughout the plant. Inside and outside type walland desk-mounted stations are used to communicate between roaming personnel and fixedwork locations.
Plant-wide instructions are issued using the paging feature.Internal Telephone SystemThe McGuire Nuclear Station PBX telephone system provides communication capability between telephone stations located within the plant by dialing the four-digit telephone stationcode.On-site Radio SystemRadio systems can be used for communication among operators, off-site monitoring teams,the control room, TSC and EOF.DEMNETDEMNET is the primary means of offsite communication.
This circuit allowsintercommunication among the EOF, TSC, control room, counties, and states. DEMNEToperates as an internet based (VoIP) communications system with a satellite back-up.
Shouldthe internet transfer rate become slow or unavailable, the DEMNET will automatically transferto satellite mode.Offsite Radio SystemA dedicated radio network can be used for communication with county and state warningpoints.Commercial Telephones Commercial telephone lines, which supply public telephone communications, are employed byDuke Energy. The local service provider provides primary and secondary power for their linesat the Central Office.NRC Emergency Telephone SystemThe NRC uses a Duke Energy dedicated telephone line which allows direct telephone communications from the plant to NRC regional and national offices.
The Duke Energycommunications line provides a link independent of the local public telephone network.Telephones connected to this network are located in the McGuire Control Room, Technical Support Center, and Emergency Operations Facility and can be used to establish NRCEmergency Notification System (ENS) and Health Physics Network (HPN) capability.
Document No. I Rev. 0 Page 197 of 272 ATTACHMENT 1EAL BasesThis EAL is the hot condition equivalent of the cold condition EAL CU5.1.NEI 99-01 Basis:This IC addresses a significant loss of on-site or offsite communications capabilities.
Whilenot a direct challenge to plant or personnel safety, this event warrants prompt notifications toOROs and the NRC.This IC should be assessed only when extraordinary means are being utilized to makecommunications 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 tooffsite locations, etc.).EAL-#4-The first EAL condition addresses a total loss of the communications methods used insupport of routine plant operations.
EAL-#2The second EAL condition addresses a total loss of the communications methods usedto notify all OROs of an emergency declaration.
The OROs referred to here are-(see Notcs) the State, Gaston, Catawba,
- Iredell, Lincoln, Cabarrus and Mecklenburg County EOCs7EAL-#3The third EAL addresses a total loss of the communications methods used to notify theNRC of an emergency declaration.
MNS Basis Reference(s):
- 1. MNS Emergency Plan Section F Emergency Communications
- 2. MNS Emergency Plan Section B On-Site Emergency Organization.
- 3. NEI 99-01 CU5I Document No. Rev. 0 Page 198 of 272 1 ATTACHMENT 1EAL BasesCategory:
Subcategory:
S -System Malfunction 8 -Containment FailureInitiating Condition:
Failure to isolate containment or loss of containment pressure control.EAL:SU8.1 Unusual EventEITHER:Any penetration is not isolated within 15 min. of a VALID containment isolation signalORContainment pressure
> 3 psig with EITHER a failure of both trains of NS OR failure ofboth trains of VX-CARF for -> 15 min.(Note 1)Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Mode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
VALID -An indication, report, or condition, is considered to be valid when it is verified by (1) aninstrument channel check, or (2) indications on related or redundant indicators, or (3) by directobservation by plant personnel, such that doubt related to the indicator's operability, thecondition's existence, or the report's accuracy is removed.
Implicit in this definition is the needfor timely assessment.
MNS Basis:The containment Phase B pressure setpoint (3 psig, ref. 1, 2) is the pressure at which thecontainment cooling systems should actuate and begin performing their function.
One full train of containment cooling operating per design is considered (ref. 1, 2):" One train of Containment Air Return Fan System (VX-CARF),
and" One train of Containment Spray System (NS)Once the Residual Heat Removal system is taking suction from the containment sump, withcontainment pressure greater than 3 psig and procedural
- guidance, one train of containment spray is manually aligned to the containment sump. If unable to place one NS train in serviceor without an operating train of VX-CARF (the CARF with a 10-minute delay) within 15 minutesthis EAL has been exceeded.
At this point a significant portion of the ice in the ice condenser would have melted and the NS system would be needed for containment pressure control.Document No. Rev. 0 Page 199 of 272 ATTACHMENT 1EAL BasesThe Unusual Event threshold applies after automatic or manual alignment of the containment spray system has been attempted with containment pressure greater than 3 psig and less thanone full train of NS is operating for greater than or equal to 15 minutes.The Unusual Event threshold also applies if containment pressure is greater than 3 psig and atleast one train of VX-CARF is not operating after a 10 minute delay for greater than or equal to15 minutes.
Without a single train of VX-CARF in service following actuation, the UnusualEvent should be declared regardless of whether ECCS is in injection or sump recirculation mode after 15 minutes.NEI 99-01 Basis:This addresses a failure of one or more containment penetrations to automatically isolate (close) when required by an actuation signal. It also addresses an event that results inhigh containment pressure with a concurrent failure of containment pressure control systems.Absent challenges to another fission product barrier, either condition represents potential degradation of the level of safety of the plant.For EAL-#4-the first condition, the containment isolation signal must be generated as the resulton an off-normal/accident condition (e.g., a safety injection or high containment pressure);
afailure resulting from testing or maintenance does not warrant classification.
Thedetermination of containment and penetration status -isolated or not isolated
-should bemade in accordance with the appropriate criteria contained in the plant AOPs and EOPs. The15-minute criterion is included to allow operators time to manually isolate the requiredpenetrations, if possible.
EAL4-#The second condition addresses a condition where containment pressure is greaterthan the setpoint at which containment energy (heat) removal systems are designed toautomatically
- actuate, and less than one full train of equipment is capable of operating perdesign. The 15-minute criterion is included to allow operators time to manually startequipment that may not have automatically
- started, if possible.
The inability to start therequired 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 would escalate to a Site Area Emergency in accordance with IC FS1 if there were aconcurrent loss or potential loss of either the Fuel Clad or RGSNCS fission product barriers.
MNS Basis Reference(s):
- 1. MNS Technical Specification 3.6.62. MNS Technical Specification 3.6.6 Bases3. MNS Technical Specification 3.3.24. UFSAR Section 6.2 Containment Systems5. NEI 99-01 SU7Document No. Rev. 0 Page 200 of 272 ATTACHMENT 1EAL BasesCategory:
Subcategory:
S -System Malfunction 9 -Hazardous Event Affecting Safety SystemsInitiating Condition:
Hazardous event affecting a SAFETY SYSTEM needed for the currentoperating modeEAL:SA9.1 AlertThe occurrence of any Table S-5 hazardous eventAND EITHER:" Event damage has caused indications of degraded performance in at least onetrain of a SAFETY SYSTEM needed for the current operating mode" The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component orstructure needed for the current operating mode[ Table S-5 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 ManagerMode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
EXPLOSION
-A rapid, violent and catastrophic failure of a piece of equipment due tocombustion, chemical reaction or overpressurization.
A release of steam (from high energylines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion.
Such events 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 drivebelts or overheated electrical equipment do not constitute fires. Observation of flame ispreferred but is NOT required if large quantities of smoke and heat are observed.
Document No. I Rev. 0 Page 201 of 272 ATTACHMENT 1EAL BasesFLOODING
-A condition where water is entering a room or area faster than installed equipment is capable of removal, resulting in a rise of water level within the room or area.SAFETY SYSTEM -A system required for safe plant operation, cooling down the plant and/orplacing it in the cold shutdown condition, including the ECCS. These are typically systemsclassified as safety-related (as defined in 10CFR50.2):
Those structures, systems and components that are relied upon to remain functional duringand following design basis events to assure:(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; (3) The capability to prevent or mitigate the consequences of accidents which couldresult in potential offsite exposures.
VISIBLE DAMAGE -Damage to a component or structure that is readily observable withoutmeasurements,
- testing, or analysis.
The visual impact of the damage is sufficient to causeconcern regarding the operability or reliability of the affected component or structure.
MNS Basis:* The significance of seismic events are discussed under EAL HU2.1 (ref. 1).* Internal FLOODING may be caused by events such as component
- failures, equipment misalignment, or outage activity mishaps (ref. 2).* External flooding may be due to high lake level. MNS powerhouse yard elevation is 760 ftMSL. The administration building and yard are elevation 747 ft MSL. The maximum waterlevel elevation at the site is 760.375 ft MFL (ref. 3, 4)." Seismic Category I structures are analyzed to withstand a sustained, design wind velocityof 95 mph. (ref. 5).* Areas containing functions and systems required for safe shutdown of the plant areidentified by fire area in the fire response procedure (ref. 5)." An explosion that degrades the performance of a SAFETY SYSTEM train or visiblydamages a SAFETY SYSTEM component or structure would be classified under this EAL.I Document No. I Rev. 0 1 Page 202 of 272 ATTACHMENT 1EAL BasesNEI 99-01 Basis:This IC addresses a hazardous event that causes damage to a SAFETY SYSTEM, or astructure 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 productbarrier, and therefore represents an actual or potential substantial degradation of the level ofsafety of the plant.EAL-1-h.4The first condition addresses damage to a SAFETY SYSTEM train that is inservice/operation since indications for it will be readily available.
The indications of degradedperformance should be significant enough to cause concern regarding the operability orreliability of the SAFETY SYSTEM train.EAL--1-7b4The second condition addresses damage to a SAFETY SYSTEM component that isnot in service/operation or readily apparent through indications alone, or to a structure containing SAFETY SYSTEM components.
Operators will make this determination based onthe totality of available event and damage report information.
This is intended to be a briefassessment not requiring lengthy analysis or quantification of the damage.Escalation of the emergency classification level would be via IC FS1 or AS4-RSI.MNS Basis Reference(s):
- 1. RP/OIAI5700/007 Earthquake
- 2. AP/O/AI5500/030 Plant Flooding3. UFSAR Section 2.1 Site Location4 UFSAR Section 3.4 Water Level (Flood) DesignI5.6.UFSAR Section 3.3.1 Wind LoadingsAP/O/AI5500/45 Plant Fire7. NEI 99-01 CA6I Document No. I Rev. 0 Page 203 of 272 1 ATTACHMENT 1EAL BasesCategory E -Independent Spent Fuel Storage Installation (ISFSI)EAL Group: ANY (EALs in this category are applicable to anyplant condition, hot or cold)An independent spent fuel storage installation (ISFSI) is a complex that is designed andconstructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.
A significant amount of the radioactive material contained within a cask/canister must escape its packaging and enter the biosphere for there to be asignificant environmental effect resulting from an accident involving the dry storage of spentnuclear fuel. Formal offsite planning is not required because the postulated worst-case accident involving an ISFSI has insignificant consequences to the public health and safety.An Unusual Event is declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated.
A hostile security event that leads to a potential loss in the level of safety of the ISFSI is aclassifiable event under Security category EAL HS1.1.Minor surface damage that does not affect storage cask/canister boundary is excluded fromthe scope of these EALs.I Document No. I Rev. 0 1 Page 204 of 272I ATTACHMENT 1EAL BasesCategory:
Sub-category:
E -ISFSINoneInitiating Condition:
Damage to a loaded cask CONFINEMENT BOUNDARYEAL:EUI.1 Unusual EventDamage to a loaded canister CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading on the surface of a loaded spent fuel cask > any Table E-1 dose limitTable E-1 ISFSI Dose LimitsNAC Magnastor NAC UMS Transnuclear (TN-32)* 190 mrem/hr 0 100 mrem/hr
- 120 mrem/hr (gamma) or 20 mrem/hr(gamma) on the side (neutron
+ (neutron) on top of the caskof the cask (excludes gamma) on theair inlet/outlet ports) side of the cask *340 mrem/hr (gamma) or 40 mrem/hr(neutron) on the sides of the radial neutron* 10 mrem/hr a 100 mrem/hr shield(neutron) on the side (neutron
+ 0 560 mrem/hr (gamma) or 280 mrem/hrof the cask (excludes gamma) on the top (neutron) on the side surfaces above theair inlet/outlet ports) of the cask radial neutron shield region*900 mrem/hr *200 mrem/hr(neutro 900 gamm a)200 (neutro 220 mrem/hr (gamma) or 400 mrem/hron the top of the gamma) at air (neutron) on the side surfaces below thecask (excludes air inlets and outlets radial neutron shield regioninlet/outlet ports)Mode Applicability:
AllDefinition(s):
CONFINEMENT BOUNDARY-The barrier(s) between spent fuel and the environment oncethe spent fuel is processed for dry storage.
As related to the MNS ISFSI, Confinement Boundary is defined as the Transportable Storage Canister (TSC) for TN, UMS andMAGNASTOR storage systems.Document No. Rev. 0 Page 205 of 272]
ATTACHMENT 1EAL BasesMNS Basis:The MNS ISFSI utilizes three designs for dry spent fuel storage:" The Transnuclear (TN) TN-32 dry spent fuel storage system" The NAC-UMS dry spent fuel storage system" The NAC-MAGNASTOR dry spent fuel storage systemAll systems consist of a Transportable Storage Canister (TSC) and concrete Vertical ConcreteCask (VCC). The TSC is the CONFINEMENT BOUNDARY for all systems.
The TSC iswelded/bolted and designed to provide confinement of all radionuclides under normal, off-normal, and accident conditions (ref. 1, 2, 3).Confinement boundary is defined as the barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage.
Therefore, damage to a confinement boundary must be a confirmed physical breach between the spent fuel and the environment for the TSC.The values shown in Table E-1 represent 2 times the limits specified in the ISFSI Certificate ofCompliance Technical Specification for radiation external to a loaded cask for each of theNAC-MAGNASTOR, NAC-UMS and TN designs.
All Table E-1 ISFSI dose limits are based onsurveys taken consistent with the locations specified in the associated Technical Specification (ref. 1,2, 3).NEI 99-01 Basis:This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of astorage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storagebeginning at the point that the loaded storage cask is sealed. The issues of concern are thecreation of a potential or actual release path to the environment, degradation of one or morefuel assemblies due to environmental
- factors, and configuration changes which could causechallenges 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 RAU1, is used here todistinguish between non-emergency and emergency conditions.
The emphasis for thisclassification is the degradation in the level of safety of the spent fuel cask and not themagnitude of the associated dose or dose rate. It is recognized that in the case of extremedamage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may bedetermined based on measurement of a dose rate at some distance from the cask.Security-related events for ISFSIs are covered under ICs HU1 and HA1.IDocument No. I Rev. 0 Page 206 of 272]
ATTACHMENT 1EAL BasesMNS Basis Reference(s):
- 1. TN Generic Technical Specifications
- 2. NAC-UMS Certificate of Compliance
- 3. MAGNASTOR Technical Specifications and Design Features4. NEI 99-01 E-HU1Document No. Rev. 0 Page 207 of 272]
ATTACHMENT 1EAL BasesCategory F -Fission Product Barrier Degradation EAL Group: Hot Conditions (NCS temperature
> 2000F); EALs inthis category are applicable only in one or more hotoperating modes.EALs in this category represent threats to the defense in depth design concept that precludes the release of highly radioactive fission products to the environment.
This concept relies onmultiple physical barriers any one of which, if maintained intact, precludes the release ofsignificant amounts of radioactive fission products to the environment.
The primary fissionproduct barriers are:A. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains thefuel pellets.B. Reactor Coolant System (NCS): The NCS Barrier includes the NCS primary side and itsconnections up to and including the pressurizer safety and relief valves, and otherconnections up to and including the primary isolation valves.C. Containment (CMT): The Containment Barrier includes the containment building andconnections up to and including the outermost containment isolation valves. This barrieralso includes the main steam, feedwater, and blowdown line extensions outside thecontainment building up to and including the outermost secondary side isolation valve.Containment Barrier thresholds are used as criteria for escalation of the ECL from Alertto a Site Area Emergency or a General Emergency.
The EALs in this category require evaluation of the loss and potential loss thresholds listed inthe fission product barrier matrix of Table F-1 (Attachment 2). "Loss" and "Potential Loss"signify the relative damage and threat of damage to the barrier.
"Loss" means the barrier nolonger assures containment of radioactive materials.
"Potential Loss" means integrity of thebarrier is threatened and could be lost if conditions continue to degrade.
The number ofbarriers that are lost or potentially lost and the following criteria determine the appropriate emergency classification level:Alert:Any loss or any potential loss of either Fuel Clad or NCSSite Area Emergency:
Loss or potential loss of any two barriersGeneral Emergency:
Loss of any two barriers and loss or potential loss of third barrierThe logic used for emergency classification based on fission product barrier monitoring shouldreflect the following considerations:
- The Fuel Clad Barrier and the NCS Barrier are weighted more heavily than theContainment Barrier.Document No. Rev. 0 Page 208 of 272 ATTACHMENT 1EAL Bases* Unusual Event ICs associated with NCS and Fuel Clad Barriers are addressed underSystem Malfunction ICs." For accident conditions involving a radiological
- release, evaluation of the fission productbarrier thresholds will need to be performed in conjunction with dose assessments toensure correct and timely escalation of the emergency classification.
For example, anevaluation of the fission product barrier thresholds may result in a Site Area Emergency classification while a dose assessment may indicate that an EAL for GeneralEmergency IC RG1 has been exceeded.
" The fission product barrier thresholds specified within a scheme reflect plant-specific MNS design and operating characteristics.
" As used in this category, the term NCS leakage encompasses not just those typesdefined in Technical Specifications but also includes the loss of NCS mass to anylocation-inside the primary containment, an interfacing system, or outside of theprimary containment.
The release of liquid or steam mass from the NCS due to the as-designed/expected operation of a relief valve is not considered to be NCS leakage." At the Site Area Emergency level, EAL users should maintain cognizance of how farpresent conditions are from meeting a threshold that would require a GeneralEmergency declaration.
For example, if the Fuel Clad and NCS fission product barrierswere both lost, then there should be frequent assessments of containment radioactive inventory and integrity.
Alternatively, if both the Fuel Clad and NCS fission productbarriers were potentially lost, the Emergency Coordinator/EOF Director would havemore assurance that there was no immediate need to escalate to a GeneralEmergency.
I Document No. Rev. 0 Page 209 of 272 1 ATTACHMENT 1EAL BasesCategory:
Subcategory:
Fission Product Barrier Degradation N/AInitiating Condition:
Any loss or any potential loss of either Fuel Clad or NCSEAL:FAI.1 AlertAny loss OR any potential loss of either Fuel Clad or NCS (Table F-1)Mode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
NoneMNS Basis:Fuel Clad, NCS and Containment comprise the fission product barriers.
Table F-1 (Attachment
- 2) lists the fission product barrier thresholds, bases and references.
At the Alert classification level, Fuel Clad and NCS barriers are weighted more heavily thanthe Containment barrier.
Unlike the Containment
- barrier, loss or potential loss of either theFuel Clad or NCS barrier may result in the relocation of radioactive materials or degradation ofcore cooling capability.
Note that the loss or potential loss of Containment barrier incombination with loss or potential loss of either Fuel Clad or NCS barrier results in declaration of a Site Area Emergency under EAL FS1.1NEI 99-01 Basis:NoneMNS Basis Reference(s):
- 1. NEI 99-01 FA1IDocument No. I Rev. 0 1 Page 210 of 272]
ATTACHMENT 1EAL BasesCategory:
Subcategory:
Fission Product Barrier Degradation N/AInitiating Condition:
Loss or potential loss of any two barriersEAL:FSI.1 Site Area Emergency Loss OR potential loss of any two barriers (Table F-I)Mode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
NoneMNS Basis:Fuel Clad, NCS and Containment comprise the fission product barriers.
Table F-1 (Attachment
- 2) lists the fission product barrier thresholds, bases and references.
At the Site Area Emergency classification level, each barrier is weighted equally.
A Site AreaEmergency is therefore appropriate for any combination of the following conditions:
" One barrier loss and a second barrier loss (i.e., loss -loss)* One barrier loss and a second barrier potential loss (i.e., loss -potential loss)* One barrier potential loss and a second barrier potential loss (i.e., potential loss -potential loss)At the Site Area Emergency classification level, the ability to dynamically assess the proximity of present conditions with respect to the threshold for a General Emergency is important.
Forexample, the existence of Fuel Clad and NCS Barrier loss thresholds in addition to offsite doseassessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification.
Alternatively, if bothFuel Clad and NCS potential loss thresholds
- existed, the Emergency Coordinator/EOF Director would have greater assurance that escalation to a General Emergency is lessimminent.
NEI 99-01 Basis:NoneDocument No. Rev. 0 Page 211 of 272 ATTACHMENT 1EAL BasesMNS Basis Reference(s):
- 1. NEI 99-01 FS1I Document No. I Rev. 0 Page 212 of 272 1 ATTACHMENT 1EAL BasesCategory:
Fission Product Barrier Degradation Subcategory:
N/AInitiating Condition:
Loss of any two barriers and loss or potential loss of third barrierEAL:FGI.1 General Emergency Loss of any two barriersANDLoss OR potential loss of third barrier (Table F-1)Mode Applicability:
1 -Power Operations, 2 -Startup, 3 -Hot Standby, 4 -Hot ShutdownDefinition(s):
NoneMNS Basis:Fuel Clad, NCS and Containment comprise the fission product barriers.
Table F-1 (Attachment
- 2) lists the fission product barrier thresholds, bases and references.
At the General Emergency classification level each barrier is weighted equally.
A GeneralEmergency is therefore appropriate for any combination of the following conditions:
- Loss of Fuel Clad, NCS and Containment barriers" Loss of Fuel Clad and NCS barriers with potential loss of Containment barrier" Loss of NCS and Containment barriers with potential loss of Fuel Clad barrier" Loss of Fuel Clad and Containment barriers with potential loss of NCS barrierNEI 99-01 Basis:NoneMNS Basis Reference(s):
- 1. NEI 99-01 FG1I Document No. Rev. 0 Page 213 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesIntroduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the threefission product barriers (Fuel Clad, Reactor Coolant System, and Containment).
The table isstructured so that each of the three barriers occupies adjacent columns.
Each fission productbarrier column is further divided into two columns; one for Loss thresholds and one forPotential Loss thresholds.
The first column of the table (to the left of the Fuel Clad Loss column) lists the categories (types) of fission product barrier thresholds.
The fission product barrier categories are:A. NCS or SG Tube LeakageB. Inadequate Heat removalC. CMT Radiation
/ NCS ActivityD. CMT Integrity or BypassE. Emergency Coordinator JudgmentEach category occupies a row in Table F-1 thus forming a matrix defined by the categories.
The intersection of each row with each Loss/Potential Loss column forms a cell in which oneor more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold fora barrier Loss/Potential Loss, the word "None" is entered in the cell.Thresholds are assigned sequential numbers within each Loss and Potential Loss columnbeginning with number one. In this manner, a threshold can be identified by its category titleand number. For example, the first Fuel Clad barrier Loss in Category A would be assigned"FC Loss A.1," the third Containment barrier Potential Loss in Category C would be assigned"CMT P-Loss C.3," etc.If a cell in Table F-1 contains more than one numbered threshold, each of the numberedthresholds, if exceeded, signifies a Loss or Potential Loss of the barrier.
It is not necessary toexceed all of the thresholds in a category before declaring a barrier Loss/Potential Loss.Subdivision of Table F-1 by category facilitates association of plant conditions to theapplicable fission product barrier Loss and Potential Loss thresholds.
This structure promotesa systematic approach to assessing the classification status of the fission product barriers.
When equipped with knowledge of plant conditions related to the fission product barriers, theEAL-user first scans down the category column of Table F-I, locates the likely category andthen reads across the fission product barrier Loss and Potential Loss thresholds in thatcategory to determine if a threshold has been exceeded.
If a threshold has not beenexceeded, the EAL-user proceeds to the next likely category and continues review of thethresholds in the new categoryIf the EAL-user determines that any threshold has been exceeded, by definition, the barrier islost or potentially lost -even if multiple thresholds in the same barrier column are exceeded, only that one barrier is lost or potentially lost. The EAL-user must examine each of the threeDocument No. Rev. 0 1 Page 214 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and Basesfission product barriers to determine if other barrier thresholds in the category are lost orpotentially lost. For example, if containment radiation is sufficiently high, a Loss of the FuelClad and NCS barriers and a Potential Loss of the Containment barrier can occur. BarrierLosses and Potential Losses are then applied to the algorithms given in EALs FG1.1, FS1.1,and FA1.1 to determine the appropriate emergency classification.
In the remainder of this Attachment, the Fuel Clad barrier threshold bases appear first,followed by the NCS barrier and finally the Containment barrier threshold bases. In eachbarrier, the bases are given according category Loss followed by category Potential Lossbeginning with Category A, then B,..., E.I Document No. Rev. 0 Page 215 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesTable F-1 Fission Product Barrier Threshold MatrixFuel Clad (FC) Barrier Reactor Coolant System (NCS) Barrier Containment (CMT) BarrierCategory Loss Potential Loss Loss Potential Loss Loss Potential Loss1. An automatic or manual ECCS 1. Operation of a standby chargingA (SI) actuation required by pump is required by EITHER:EITHER: -UNISOLABLE NCS leakage 1. A leaking or RUPTURED SG iso UNISOLABLE NCS -SG tube leakage FAULTED outside of containment SG Tube leakageLeakage 2. Integrity-RED PATH conditions
" SG tube RUPTURE met1. Core Cooling-ORANGE PATHconditions met 1. Heat Sink-RED PATH conditions
- 1. Core Coaling-RED PATH1 Core Cooling-RED PATH 2. Heat Sink-RED PATH conditions metInadequate conditions met met None AND None ANDAND Heat sink is required Restoration procedures notRemoval Heat sink is required effective within 15 min. (Note 1)Heat sink is requiredC 1. EMF51AIB
>Table F-2 column 1. EMF51AJB
> Table F-2 columnCMT "FC Loss" None "NCS Loss" None None 1. EMF51 AB > Table F-2 columnRadiation
- 2. Dose equivalent 1-131 coolant "CMT Potential Loss"I NCS activity
> 300 pCi/gmActivity1. Containment isolation isrequiredAND EITHER: 1. Containment-RED Path conditions
-Containment integrity has metbeen lost based on 2. Containment hydrogen concentratior Emergency
> 6%CMT None None None None Coordinator/EOF Directorjudgment
- 3. Containment pressure
> 3 psigwith EITHER a failure of bothor Bypass
- UNISOLABLE pathway from trains of NS OR failure of bothContainment to the environment trains of VX-CARF for > 15 min.exists (Note 1)2. Indications of NCS leakageoutside of containment E 1. Any condition in the opinion of 1. Any condition in the opinion of 1. Any condition in the opinion of 1. Any condition in the opinion of the 1. Any condition in the opinion of 1. Any condition in the opinion of thethe Emergency the Emergency Coordinator/EOF the Emergency Coordinator/EOF Emergency Coordinator/EOF the Emergency Coordinator/EOF Emergency Coordinator/EOF EC Coordinator/EOF Director that Director that indicates potential Director that indicates loss of the Director that indicates potential Director that indicates loss of the Director that indicates potential ud en barer loss of the fuel clad barrier NCS barrier loss of the NCS barrier containment barrier loss of the containment barrierJudgment barrierI Document No. I Rev. 0 1 Page 216 of 272 I ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Fuel CladCategory:
NoneI Document No. Rev. 0 Page 217 of 272]
ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Fuel CladCategory:
NoneIDocument No. I Rev. 0 1 Page 218 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFuel CladBarrier:Category:
B. Inadequate Heat RemovalDegradation Threat: LossThreshold:
- 1. Core Cooling-RED PATH conditions metDefinition(s):
NoneBasis:Plant-Specific Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery.
The CSFSTs are normally monitored using theSPDS display on the Operator Aid Computer (OAC) (ref. 1).GenericThis reading indicates temperatures within the core are sufficient to cause significant superheating of reactor coolant.MNS Basis Reference(s):
- 1. EP/1 (2)/A/5000/F-0 Critical Safety Function Status Trees2. EP/1(2)/N5000/FR-C.1 Response to Inadequate Core Cooling3. EP/I(2)/IA5000/FR-C.2 Response to Degraded Core Cooling4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.AI Document No. I Rev. 0 Page 219 of 272 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFuel CladBarrier:Category:
B. Inadequate Heat RemovalDegradation Threat: Potential LossThreshold:
- 1. Core Coln-RNePath conditions metDefinition(s):
NoneBasis:Plant-Specific Critical Safety Function Status Tree (CSFST) Core Cooling-ORANGE path indicates indicates subcooling has been lost and that some fuel clad damage may potentially occur. The CSFSTsare normally monitored using the SPDS display on the Operator Aid Computer (OAC) (ref. 1).GenericThis reading indicates a reduction in reactor vessel water level sufficient to allow the onset ofheat-induced cladding damage.MNS Basis Reference(s):
- 1. EP/1(2)/A/5000/F-0 Critical Safety Function Status Trees2. EP/1(2)/A/5000/FR-C.1 Response to Inadequate Core Cooling3. EP/I(2)/IA5000/FR-C.2 Response to Degraded Core Cooling4. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.AI Document No. Rev. 0 1 Page 220 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Fuel CladCategory:
B. Inadequate Heat RemovalDegradation Threat: Potential LossThreshold:
- 2. Heat Sink-RED Path conditions metANDHeat sink is requiredDefinition(s):
NoneBasis:Plant-Specific In combination with NCS Potential Loss B.1, meeting this threshold results in a Site AreaEmergency.
Critical Safety Function Status Tree (CSFST) Heat Sink-RED path indicates the ultimate heatsink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer(OAC) (ref. 1).The phrase "and heat sink required" precludes the need for classification for conditions inwhich NCS pressure is less than SG pressure or Heat Sink-RED path entry was createdthrough operator action directed by an EOP. For example, FR-H.1 is entered from CSFSTHeat Sink-Red.
Step 2 tells the operator to determine if heat sink is required by checking thatNCS pressure is greater than any non-faulted SG pressure and NCS Thot is greater than 350°F(3470F ACC). If these conditions exist, Heat Sink is required.
Otherwise, the operator is toeither return to the procedure and step in effect or place ND in service for heat removal.
Forlarge LOCA events inside the Containment, the SGs are moot because heat removal throughthe containment heat removal systems takes place. Therefore, Heat Sink Red should not berequired and, should not be assessed for EAL classification because a LOCA event aloneshould not require higher than an Alert classification.
(ref. 2).GenericThis condition indicates an extreme challenge to the ability to remove RGS-NCS heat usingthe steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the Fuel Clad Barrier.
In accordance with EOPs, there may beDocument No. I Rev. 0 1 Page 221 of 272 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and Basesunusual accident conditions during which operators intentionally reduce the heat removalcapability of the steam generators; during these conditions, classification using threshold is notwarranted.
MNS Basis Reference(s):
- 1. EP/1 (2)/A/5000/F-0 Critical Safety Function Status Trees2. EP/I1(2)/A/5000/FR-H.1 Response to Loss of Secondary Heat Sink3. NEI 99-01 Inadequate Heat Removal Fuel Clad Loss 2.BI Document No. Rev. 0 Page 222 of 272 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFuel CladBarrier:Category:
C. CMT Radiation
/ NCS ActivityDegradation Threat: LossThreshold:
- 1. EMF51A/B
> Table F-2 column "FC Loss"Table F-2 Containment Radiation
-R/hr (EMF51A & B)Time After S/D (Hrs.) NCS Loss FC Loss CMT Potential Loss0-1 8.8 550 55001-2 8.4 400 40002-8 7.0 160 1600>8 6.2 100 1000Definition(s):
NoneBasis:Plant-Specific The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) ismonitored by the containment high range monitors, EMF51A & B. EMF51 & B are locatedinside containment.
The detector range is approximately 1 to 1 E8 R/hr (logarithmic scale).Radiation Monitors EMF51A & B provide a diverse means of measuring the containment forhigh level gamma radiation.
(ref. 1).The Table F-2 values, column FC Loss represents, based on core damage assessment procedure, the expected containment high range radiation monitor (EMF51A & B) responsebased on a LOCA, for periods of 1, 2, 8 and >8 hours after shutdown, no sprays and NCSpressure
< 1600 psig with -2% fuel failure (ref. 1).The value is derived as follows:RP/O/A/5700/019 Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage 1, 2,and 8 and >8 hours after shutdown without spray and NCS pressure
< 1600 psig x 0.02(rounded)
(ref. 1).Document No. Rev. 0 Page 223 of 272]
ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesGenericThe radiation monitor reading corresponds to an instantaneous release of all reactor coolantmass into the containment, assuming that reactor coolant activity equals 300 pCi/gm doseequivalent 1-131. Reactor coolant activity above this level is greater than that expected foriodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Sincethis condition indicates that a significant amount of fuel clad damage has occurred, itrepresents a loss of the Fuel Clad Barrier.The radiation monitor reading in this threshold is higher than that specified for R-,S-NCSBarrier Loss threshold
-AC.1 since it indicates a loss of both the Fuel Clad Barrier and theRGS-NCS Barrier.
Note that a combination of the two monitor readings appropriately escalates the omo.-goncy clacc!ficetion IYooECL to a Site Area Emergency.
MNS Basis Reference(s):
- 1. RP/O/A/5700/019 Core Damage Assessment
- 2. NEI 99-01 CMT Radiation
/ RCS Activity Fuel Clad Loss 3.AI Document No. I Rev. 0 Page 224 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFuel CladBarrier:Category:
C. CMT Radiation
/ NCS ActivityDegradation Threat: LossThreshold:
- 2. Dose equivalent 1-131 coolant activity
> 300 pCi/gmDefinition(s):
NoneBasis:Plant-Specific Elevated reactor coolant activity represents a potential degradation in the level of safety of theplant and a potential precursor of more serious problems.
The threshold dose equivalent 1-131concentration is well above that expected for iodine spikes and corresponds to about 2% fuelclad damage. When reactor coolant activity reaches this level the Fuel Clad barrier isconsidered lost. (ref. 1).GenericThis threshold indicates that RGS NCS radioactivity concentration is greater than 300 pCi/gmdose equivalent 1-131. Reactor coolant activity above this level is greater than that expectedfor 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, itrepresents a loss of the Fuel Clad Barrier.There is no Potential Loss threshold associated with RG, NCS Activity
/ Containment Radiation.
MNS Basis Reference(s):
- 1. RP/O/A/5700/019 Core Damage Assessment
- 2. NEI 99-01 CMT Radiation
/ RCS Activity Fuel Clad Loss 3.BIII Document No. Rev. 0 1 Page 225 of 272 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Fuel CladCategory:
C. CMT Radiation
/ NCS ActivityDegradation Threat: Potential LossThreshold:
NoneDocument No. Rev. 0 Page 226 of 272]
ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Fuel CladCategory:
D. CMT Integrity or BypassDegradation Threat: LossThreshold:
NoneDocument No. Rev. 0 Page 227 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Fuel CladCategory:
D. CMT Integrity or BypassDegradation Threat: Potential LossThreshold:
NoneDocument No. Rev. 0 Page 228 of 272:]
ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Fuel CladCategory:
E. Emergency Coordinator JudgmentDegradation Threat: LossThreshold:
- 1. Any condition in the opinion of the Emergency Coordinator/EOF Director that indicates loss of the Fuel Clad barrierDefinition(s):
NoneBasis:Plant-Specific The Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the Fuel Clad barrier is lost. Such a determination should include imminentbarrier degradation, barrier monitoring capability and dominant accident sequences.
" Imminent barrier degradation exists if the degradation will likely occur within two hoursbased on a projection of current safety system performance.
The term "imminent" refersto recognition of the inability to reach safety acceptance criteria before completion of allchecks." Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability
- concerns, readings fromportable instrumentation and consideration of offsite monitoring results." Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator/EOF Director should be mindfulof the Loss of AC power (Station Blackout) and ATWS EALs to assure timelyemergency classification declarations.
GenericThis threshold addresses any other factors that are to be used by the Emergency QireGterCoordinator in determining whether the Fuel Clad barrier is lostMNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.ADocument No. I Rev. 0 Page 229 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Fuel CladCategory:
E. Emergency Coordinator JudgmentDegradation Threat: Potential LossThreshold:
- 1. Any condition in the opinion of the Emergency Coordinator/EOF Director that indicates potential loss of the Fuel Clad barrierBasis:Plant-Specific The Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the Fuel Clad barrier is potentially lost. Such a determination should includeimminent barrier degradation, barrier monitoring capability and dominant accident sequences.
- Imminent barrier degradation exists if the degradation will likely occur within two hoursbased on a projection of current safety system performance.
The term "imminent" refersto recognition of the inability to reach safety acceptance criteria before completion of allchecks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability
- concerns, readings fromportable instrumentation and consideration of offsite monitoring results." Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
GenericThis threshold addresses any other factors that are to be used by the Emergency CoordinatorgieGtef in determining whether the Fuel Clad barrier is potentially lost. TheEmergency DiQeoter-Coordinator should also consider whether or not to declare the barrierpotentially lost in the event that barrier status cannot be monitored.
MNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.ADocument No. Rev. 0 Page 230 of 2721 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesReactor Coolant SystemBarrier:Category:
A. NCS or SG Tube LeakageDegradation Threat: LossThreshold:
- 1. An automatic or manual ECCS (SI) actuation required by EITHER:" UNISOLABLE NCS leakage" SG tube RUPTUREDefinition(s):
UNISOLABLE
-An open or breached system line that cannot be isolated, remotely or locally.RUPTURE -The condition of a steam generator in which primary-to-secondary leakage is ofsufficient magnitude to require a safety injection.
Basis:GenericECCS (SI) actuation is caused by (ref. 1):* Pressurizer pressure
< 1845 psig* Containment pressure
> 1.0 psigGenericThis threshold is based on an UNISOLABLE RGS-NCS leak of sufficient size to require anautomatic or manual actuation of the Emergency Core Cooling System (ECCS). Thiscondition clearly represents a loss of the RGS-NCS Barrier.This threshold is applicable to unidentified and pressure boundary
- leakage, as well asidentified leakage.
It is also applicable to UNISOLABLE RGS-NCS leakage through aninterfacing system. The mass loss may be into any location
-inside containment, to thesecondary-side (i.e., steam generator tube leakage) or outside of containment.
A steam generator with primary-to-secondary leakage of sufficient magnitude to require asafety injection is considered to be RUPTURED.
If a RUPTURED steam generator is alsoFAULTED outside of containment, the declaration escalates to a Site Area Emergency sincethe Containment Barrier Loss threshold 1 .A will also be met.MNS Basis Reference(s):
- 1. EP/1(2)/A/5000/E-0 Reactor Trip or Safety Injection
- 2. EP/1(2)/A/5000/E-3 Steam Generator Tube RuptureDocument No. I Rev. 0 Page 231 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and Bases3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Loss 1.AIDocument No. I Rev. 0 1 Page 232 of 272]
ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesReactor Coolant SystemBarrier:Category:
A. NCS or SG Tube LeakageDegradation Threat: Potential LossThreshold:
- 1. Operation of a standby charging pump is required by EITHER:" UNISOLABLE NCS leakage" SG tube RUPTUREDefinition(s):
UNISOLABLE
-An open or breached system line that cannot be isolated, remotely or locally.RUPTURE -The condition of a steam generator in which primary-to-secondary leakage is ofsufficient magnitude to require a safety injection.
Basis:GenericThe Chemical and Volume Control System (CVCS) includes two centrifugal charging pumpswhich take suction from the Volume Control Tank and return cooled, purified reactor coolant tothe NCS. Normal charging flow is handled by one of the two charging pumps. Each chargingpump is designed for a flow rate of 150 gpm. A second charging pump being required isindicative of a substantial NCS leak. (ref. 1).GenericThis threshold is based on an UNISOLABLE RGS-NCS leak that results in the inability tomaintain pressurizer level within specified limits by operation of a normally used charging(makeup) pump, but an ECCS (SI) actuation has not occurred.
The threshold is met when anoperating procedure, or operating crew supervision, directs that a standby charging (makeup)pump be placed in service to restore and maintain pressurizer level.This threshold is applicable to unidentified and pressure boundary
- leakage, as well asidentified leakage.
It is also applicable to UNISOLABLE RGS-NCS leakage through aninterfacing system. The mass loss may be into any location
-inside containment, to thesecondary-side (i.e., steam generator tube leakage) or outside of containment.
If a leaking steam generator is also FAULTED outside of containment, the declaration escalates to a Site Area Emergency since the Containment Barrier Loss threshold 1 .A will alsobe met.Document No. Rev. 0 Page 233 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesMNS Basis Reference(s):
- 1. UFSAR Section 9.3.4 Chemical and Volume Control System2. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss 1.AI Document No. I Rev. 0 Page 234 of 272]
ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesReactor Coolant SystemBarrier:Category:
A. NCS or SG Tube LeakageDegradation Threat: Potential LossThreshold:
- 2. Integrity-RED path conditions metDefinition(s):
NoneBasis:GenericThe "Potential Loss" threshold is defined by the CSFST Reactor Coolant Integrity
-Red Path plant conditions and associated PTS Limit A indicates anextreme challenge to the safety function when plant parameters are to the right of the limitcurve following excessive NCS cooldown under pressure (ref. 1, 2).GenericThis condition indicates an extreme challenge to the integrity of the RGS-NCS pressureboundary due to pressurized thermal shock -a transient that causes rapid RQS-NCScooldown while the RGS-NCS is in Mode 3 or higher (i.e., hot and pressurized).
MNS Basis Reference(s):
- 1. EP/1 (2)IA/5000/F-0 Critical Safety Function Status Trees2. EP/I(2)/A/5000/FR-P.1 Response to Imminent Pressurized Thermal Shock Condition
- 3. NEI 99-01 RCS or SG Tube Leakage Reactor Coolant System Potential Loss I.BDocument No. Rev. 0 1 Page 235 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Reactor Coolant SystemCategory:
B. Inadequate Heat RemovalDegradation Threat: LossThreshold:
NoneIDocument No. I Rev. 0 Page 236 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesReactor Coolant SystemBarrier:Category:
Degradation Threat:B. Inadequate Heat RemovalPotential LossThreshold:
Definition(s):
NoneBasis:Plant-Specific In combination with FC Potential Loss B.2, meeting this threshold results in a Site AreaEmergency.
Critical Safety Function Status Tree (CSFST) Heat Sink-RED path indicates the ultimate heatsink function is under extreme challenge and that some fuel clad damage may potentially occur (ref. 1).The CSFSTs are normally monitored using the SPDS display on the Operator Aid Computer(OAC) (ref. 1).The phrase "and heat sink required" precludes the need for classification for conditions inwhich NCS pressure is less than SG pressure or Heat Sink-RED path entry was createdthrough operator action directed by an EOP. For example, FR-H.1 is entered from CSFSTHeat Sink-Red.
Step 2 tells the operator to determine if heat sink is required by checking thatNCS pressure is greater than any non-faulted SG pressure and NCS Thot is greater than3500F. If these conditions exist, Heat Sink is required.
Otherwise, the operator is to eitherreturn to the procedure and step in effect or place ND in service for heat removal.
For largeLOCA events inside the Containment, the SGs are moot because heat removal through thecontainment heat removal systems takes place. Therefore, Heat Sink Red should not berequired and, should not be assessed for EAL classification because a LOCA event aloneshould not require higher than an Alert classification.
(ref. 1, 2)GenericThis condition indicates an extreme challenge to the ability to remove RGS-NCS heat usingthe steam generators (i.e., loss of an effective secondary-side heat sink). This condition represents a potential loss of the RGS-NCS Barrier.
In accordance with EOPs, there may beDocument No. I Rev. 0 1 Page 237 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and Basesunusual accident conditions during which operators intentionally reduce the heat removalcapability of the steam generators; during these conditions, classification using threshold is notwarranted.
Meeting this threshold results in a Site Area Emergency because this threshold is identical toFuel Clad Barrier Potential Loss threshold 2.B.2; both will be met. This condition warrants aSite Area Emergency declaration because inadequate RGS-NCS heat removal may result infuel heat-up sufficient to damage the cladding and increase RGS-NCS pressure to the pointwhere mass will be lost from the system.MNS Basis Reference(s):
- 1. EP/1 (2)/A/5000/F-0 Critical Safety Function Status Trees2. EP/1 (2)/A/5000/FR-H.1 Response to Loss of Secondary Heat Sink3. NEI 99-01 Inadequate Heat Removal NCS Loss 2.BI Document No. Rev. 0 1 Page 238 of 272 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesReactor Coolant SystemBarrier:Category:
C. CMT Radiation/
NCS ActivityDegradation Threat: LossThreshold:
- 1. EMF51A/B
> Table F-2 column "NCS Loss"Table F-2 Containment Radiation
-R/hr (EMF51A & B)Time After S/D (Hrs.) NCS Loss FC Loss CMT Potential Loss0-1 8.8 550 55001-2 8.4 400 40002-8 7.0 160 1600>8 6.2 100 1000Definition(s):
N/ABasis:Plant-Specific The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) ismonitored by the containment high range monitors, EMF51A & B. EMF51A & B are locatedinside containment.
The detector range is approximately 1 to 1 E8 R/hr (logarithmic scale).Radiation Monitors EMF51A & B provide a diverse means of measuring the containment forhigh level gamma radiation.
(ref. 1).The value specified represents, based on core damage assessment procedure RP/0/A/5700/019 Figure 1, the expected containment high range radiation monitor (EMF51A &B) response based on a LOCA, for periods of 1, 2, 8 and >8 hours after shutdown with no fuelfailure (ref. 1).The value is derived as follows:RP/0/A/5000/019 Figure 1 Containment Radiation Level vs. Time for NCS Release for periodsof 1, 2, 8 and >8 hours after shutdown (rounded)
(ref. 1).Document No. Rev. 0 Page 239 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesGenericThe radiation monitor reading corresponds to an instantaneous release of all reactor coolantmass into the containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Lossthreshold
,=%AC.1 since it indicates a loss of the RCS-NCS Barrier only.There is no Potential Loss threshold associated with RGS-NCS Activity
/ Containment Radiation.
MNS Basis Reference(s):
- 1. RP/O/A/5700/019 Core Damage Assessment
- 2. NEI 99-01 CMT Radiation
/ RCS Activity NCS Loss 3.ADocument No. Rev. 0 Page 240 of 272]
ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Reactor Coolant SystemCategory:
B. CMT Radiation/
NCS ActivityDegradation Threat: Potential LossThreshold:
I NoneI Document No. I Rev. 0 Page 241 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Reactor Coolant SystemCategory:
D. CMT Integrity or BypassDegradation Threat: LossThreshold:
NoneI Document No. I Rev. 0 Page 242 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Reactor Coolant SystemCategory:
D. CMT Integrity or BypassDegradation Threat: Potential LossThreshold:
NoneDocument No. Rev. 0 Page 243 of 272:]
ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Reactor Coolant SystemCategory:
E. Emergency Coordinator JudgmentDegradation Threat: LossThreshold:
- 1. Any condition in the opinion of the Emergency Coordinator/EOF Director that indicates loss of the NCS barrierDefinition(s):
NoneBasis:Plant-Specific The Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the NCS barrier is lost. Such a determination should include imminent barrierdegradation, barrier monitoring capability and dominant accident sequences.
- Imminent barrier degradation exists if the degradation will likely occur within two hoursbased on a projection of current safety system performance.
The term "imminent" refersto the recognition of the inability to reach safety acceptance criteria before completion of all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability
- concerns, readings fromportable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
GenericThis threshold addresses any other factors that may be used by the Emergency DQiecteiCoordinator in determining whether the NCS Barrier is lost.MNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment NCS Loss 6.ADocument No. Rev. 0 Page 244 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Reactor Coolant SystemCategory:
E. Emergency Coordinator JudgmentDegradation Threat: Potential LossThreshold:
- 1. Any condition in the opinion of the Emergency Coordinator/EOF Director that indicates potential loss of the NCS barrierDefinition(s):
NoneBasis:Plant-Specific The Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the NCS barrier is potentially lost. Such a determination should includeimminent barrier degradation, barrier monitoring capability and dominant accident sequences.
" Imminent barrier degradation exists if the degradation will likely occur within two hoursbased on a projection of current safety system performance.
The term "imminent" refersto the inability to reach final safety acceptance criteria before completing all checks." Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability
- concerns, readings fromportable instrumentation and consideration of offsite monitoring results.* Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
GenericThis threshold addresses any other factors that may be used by the Emergency DiieGtefCoordinator in determining whether the NCS Barrier is potentially lost. The Emergency Di§Gter-Coordinator should also consider whether or not to declare the barrier potentially lostin the event that barrier status cannot be monitored.
MNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment NCS Potential Loss 6.ADocument No. I Rev. 0 Page 245 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesContainment Barrier:Category:
A. NCS or SG Tube LeakageDegradation Threat: LossThreshold:
- 1. A leaking or RUPTURED SG is FAULTED outside of containment Definition(s):
FAULTED -The term applied to a steam generator that has a steam leak on the secondary side of sufficient size to cause an uncontrolled drop in steam generator pressure or the steamgenerator to become completely depressurized.
RUPTURED
-The condition of a steam generator in which primary-to-secondary leakage is ofsufficient magnitude to require a safety injection.
Basis:Plant-Specific None.GenericThis threshold addresses a leaking or RUPTURED Steam Generator (SG) that is alsoFAULTED outside of containment.
The condition of the SG, whether leaking or RUPTURED, is determined in accordance with the thresholds for RGSNCS Barrier Potential Loss 4-.A.1 andLoss 4-.A.1, respectively.
This condition represents a bypass of the containment barrier.FAULTED is a defined term within the NEI 99-01 methodology; this determination is notnecessarily dependent upon entry into, or diagnostic steps within, an EOP. For example, if thepressure in a steam generator is decreasing uncontrollably
([part of the FAULTED definition) and the FAULTED steam generator isolation procedure is not entered because EOP userrules are dictating implementation of another procedure to address a higher priority condition, the steam generator is still considered FAULTED for emergency classification purposes.
The FAULTED criterion establishes an appropriate lower bound on the size of a steam releasethat may require an emergency classification.
Steam releases of this size are readilyobservable with normal Control Room indications.
The lower bound for this aspect of thecontainment barrier is analogous to the lower bound criteria specified in IC SU4 for the fuelIDocument No. IRev. 0 1 Page 246 of 27 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and Basesclad barrier (i.e., RGS-NCS activity values) and IC SU5 for the RGS-NCS barrier (i.e., RGSNCS 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 auxiliary (emergency) feed waterpump. These types of conditions will result in a significant and sustained release ofradioactive steam to the environment (and are thus similar to a FAULTED condition).
Theinability to isolate the steam flow without an adverse effect on plant cooldown meets the intentof a loss of containment.
Steam releases associated with the expected operation of a SG power operated relief valve orsafety relief valve do not meet the intent of this threshold.
Such releases may occurintermittently for a short period of time following a reactor trip as operators process throughemergency operating procedures to bring the plant to a stable condition and prepare to initiatea plant cooldown.
Steam releases associated with the unexpected operation of a valve (e.g.,a stuck-open safety valve) do meet this threshold.
Following an SG tube leak or rupture, there may be minor radiological releases through asecondary-side system component (e.g., air ejectors, glad seal exhausters, valve packing,etc.). These types of releases do not constitute a loss or potential loss of containment butshould be evaluated using the Recognition Category A-R ICs.The emcrgency classification ov-olECLs resulting from primary-to-secondary
- leakage, with orwithout a steam release from the FAULTED SG, are summarized below.Affected SG is FAULTEDOutside of Containment?
P-to-S Leak RateYesNoLess than or equal to 25 gpmGreater than 25 gpmRequires operation of a standbycharging (makeup) pump (RCSNCS Barrier Potential Loss)Requires an automatic or manualECCS (SI) actuation (RCS-NCSBarrier Loss)No classification Unusual Event perSU4SU5.1Site Area Emergency perFS1.1Site Area Emergency perFS1.1No classification Unusual Event per8U4SU5.1Alert per FA1.1Alert per FA1.1There is no Potential Loss threshold associated with RGS-NCS or SG Tube Leakage.Document No. Rev. 0 Page 247 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesMNS Basis Reference(s):
- 1. EP/1(2)/A/5000/E-0 Reactor Trip or Safety Injection
- 2. EP/1(2)/A/5000/E-3 Steam Generator Tube Rupture3. NEI 99-01 RCS or SG Tube Leakage Containment Loss 1.ADocument No. Rev. 0 Page 248 of 272 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Containment
,Category:
A. NCS or SG Tube LeakageDegradation Threat: Potential LossThreshold:
NoneI Document No. Rev. 0 1 Page 249 of 272 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesContainment Barrier:Category:
B. Inadequate Heat RemovalDegradation Threat: Potential LossThreshold:
- 1. Core Cooling-RED path conditions metANDRestoration procedures not effective within 15 min. (Note 1)Definition(s):
NoneBasis:Plant-Specific Critical Safety Function Status Tree (CSFST) Core Cooling-RED path indicates significant core exit superheating and core uncovery.
The CSFSTs are normally monitored using theSPDS display on the Operator Aid Computer (OAC) (ref. 1).The function restoration procedures are those emergency operating procedures that addressthe recovery of the core cooling critical safety functions.
The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing (ref. 1, 2, 3).A direct correlation to status trees can be made if the effectiveness of the restoration procedures is also evaluated.
If core exit thermocouple (TC) readings are greater than 1,200°F(ref. 1), Fuel Clad barrier is also lost.GenericThis condition represents an IMMINENT core melt sequence which, if not corrected, couldlead to vessel failure and an increased potential for containment failure.
For this condition tooccur, there must already have been a loss of the RGSNCS Barrier and the Fuel Clad Barrier.If implementation of a procedure(s) to restore adequate core cooling is not effective (successful) within 15 minutes, it is assumed that the event trajectory will likely lead to coremelting and a subsequent challenge of the Containment Barrier.The restoration procedure is considered "effective" if core exit thermocouple readings aredecreasing and/or if reactor vessel level is increasing.
Whether or not the procedure(s) will beeffective should be apparent within 15 minutes.
The Emergency shouldescalate the emergency classification level as soon as it is determined that the procedure(s) will not be effective.
Document No. Rev. 0 Page 250 of 272I ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesSevere accident analyses (e.g., NUREG-1 150) have concluded that function restoration procedures can arrest core degradation in a significant fraction of core damage scenarios, andthat the likelihood of containment failure is very small in these events. Given this, it isappropriate to provide 15 minutes beyond the required entry point to determine if procedural actions can reverse the core melt sequence.
MNS Basis Reference(s):
- 1. EP/1(2)/A/5000/F-0 Critical Safety Function Status Trees2. EP/1(2)/A/5000/FR-C.1 Response to Inadequate Core Cooling3. EP/1(2)/A/5000/FR-C.2 Response to Degraded Core Cooling4. NEI 99-01 Inadequate Heat Removal Containment Potential Loss 2.AI Document No. I Rev. 0 Page 251 of 272 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Containment Category:
C. CMT Radiation/NCS ActivityDegradation Threat: LossThreshold:
NoneI Document No. Rev. 0 1 Page 252 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesContainment Barrier:Category:
C. CMT Radiation/NCS ActivityDegradation Threat: Potential LossThreshold:
1 .EMF5IA/B
> Table F-2 column "CMT Potential Loss"ITable F-2 Containment Radiation
-R/hr (EMF51A & B)Time After SOD (Hrs.) NCS Loss FC Loss CMT Potential Loss0-1 8.8 550 55001-2 8.4 400 40002-8 7.0 160 1600>8 6.2 100 1000Definition(s):
NoneBasis:Plant-Specific The gamma dose rate resulting from a postulated loss of coolant accident (LOCA) ismonitored by the containment high range monitors, EMF51A & B. EMF51A & B are locatedinside containment.
The detector range is approximately 1 to 1 E8 R/hr (logarithmic scale).Radiation Monitors EMF51A & B provide a diverse means of measuring the containment forhigh level gamma radiation.
(ref. 1).The Table F-2 values, column CMT Potential Loss represents, based on core damageassessment procedure, the expected containment high range radiation monitor (EMF51A & B)response based on a LOCA, for periods of 1, 2, 8 and >8 hours after shutdown, no sprays andNCS pressure
< 1600 psig with -20% fuel failure (ref. 1).The value is derived as follows:RP/0/A/5700/019 Figure 3 Containment Radiation Level vs. Time for 100% Clad Damage 1, 2,8 and >8 hours after shutdown without spray and NCS pressure
< 1600 psig x 0.20 (rounded)
(ref. 1).Document No. Rev. 0 Page 253 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesGenericThe radiation monitor reading corresponds to an instantaneous release of all reactor coolantmass into the containment, assuming that 20% of the fuel cladding has failed. This level offuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss andRCS-NCS Barrier Loss thresholds.
NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power PlantAccidents, indicates the fuel clad failure must be greater than approximately 20% in order forthere to be a major release of radioactivity requiring offsite protective actions.
For thiscondition to exist, there must already have been a loss of the RGSNCS Barrier and the FuelClad Barrier.
It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the omorgocy clacificfati.n loevoECL to a General Emergency.
MNS Basis Reference(s):
- 1. RP/O/A/5700/01 9 Core Damage Assessment
- 2. NEI 99-01 CMT Radiation
/ RCS Activity Containment Potential Loss 3.AII Document No. Rev. 0 1 Page 254 of 272 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesContainment Barrier:Category:
D. CMT Integrity or BypassDegradation Threat: LossThreshold:
- 1. Containment isolation is requiredAND EITHER:" Containment integrity has been lost based on EC judgment" UNISOLABLE pathway from containment to the environment existsDefinition(s):
UNISOLABLE
-An open or breached system line that cannot be isolated, remotely or locally.Basis:Plant-Specific NoneGenericThese thresholds address a situation where containment isolation is required and one of twoconditions exists as discussed below. Users are reminded that there may be accident andrelease conditions that simultaneously meet both bulleted thresholds 4.A.1 and 4.A.2.4A-4First Threshold
-Containment integrity has been lost, i.e., the actual containment atmospheric leak rate likely exceeds that associated with allowable leakage (or sometimes referred to as design leakage).
Following the release of RGS-NCS mass into containment, containment pressure will fluctuate based on a variety of factors; a loss of containment integrity condition may (or may not) be accompanied by a noticeable drop in containment pressure.
Recognizing the inherent difficulties in determining a containment leak rate duringaccident conditions, it is expected that the Emergency 0*eeter-Coordinator will assess thisthreshold using judgment, and with due consideration given to current plant conditions, andavailable operational and radiological data (e.g., containment
- pressure, readings on radiation monitors outside containment, operating status of containment pressure control equipment, etc.).Refer to the middle piping run of Figure 0-F-41. Two simplified examples are provided.
Oneis leakage from a penetration and the other is leakage from an in-service system valve.Depending upon radiation monitor locations and sensitivities, the leakage could be detectedby any of the four monitors depicted in the figure.Document No. Rev. 0 Page 255 of 272I ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesAnother example would be a loss or potential loss of the RGS-NCS barrier, and thesimultaneous occurrence of two FAULTED locations on a steam generator where one fault islocated inside containment (e.g., on a steam or feedwater line) and the other outside ofcontainment.
In this case, the associated steam line provides a pathway for the containment atmosphere to escape to an area outside the containment.
Following the leakage of RGS-NCS mass into containment and a rise in containment
- pressure, there may be minor radiological releases associated with allowable (design) containment leakage through various penetrations or system components.
These releases do notconstitute a loss or potential loss of containment but should be evaluated using theRecognition Category A-R ICs.4.ASecond Threshold
-Conditions are such that there is an UNISOLABLE pathway for themigration of radioactive material from the containment atmosphere to the environment.
Asused here, the term "environment" includes the atmosphere of a room or area, outside thecontainment, that may, in turn, communicate with the outside-the-plant atmosphere (e.g.,through discharge of a ventilation system or atmospheric leakage).
Depending upon a varietyof factors, this condition may or may not be accompanied by a noticeable drop in containment pressure.
Refer to the top piping run of Figure 9-F-41. In this simplified
- example, the inboard andoutboard isolation valves remained open after a containment isolation was required (i.e.,containment isolation was not successful).
There is now an UNISOLABLE pathway from thecontainment to the environment.
The existence of a filter is not considered in the threshold assessment.
Filters do not removefission product noble gases. In addition, a filter could become ineffective due to iodine and/orparticulate loading beyond design limits (i.e., retention ability has been exceeded) or watersaturation from steam/high humidity in the release stream.Leakage between two interfacing liquid systems, by itself, does not meet this threshold.
Refer to the bottom piping run of Figure -F--41. In this simplified
- example, leakage in an RGPNCP seal cooler is allowing radioactive material to enter the Auxiliary Building.
Theradioactivity would be detected by the Process Monitor.
If there is no leakage from the closedwater cooling system to the Auxiliary
- Building, then no threshold has been met. If the pumpdeveloped a leak that allowed steam/water to enter the Auxiliary
- Building, then secondthreshold-4 7-B would be met. Depending upon radiation monitor locations and sensitivities, thisleakage could be detected by any of the four monitors depicted in the figure and cause thefirst threshold 4-A-.-to be met as well.Following the leakage of RGS-NCS mass into containment and a rise in containment
- pressure, there may be minor radiological releases associated with allowable containment leakagethrough various penetrations or system components.
Minor releases may also occur if acontainment isolation valve(s) fails to close but the containment atmosphere escapes to anenclosed system. These releases do not constitute a loss or potential loss of containment butshould be evaluated using the Recognition Category A-R ICs.I Document No. I Rev. 0 Page 256 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesThe status of the containment barrier during an event involving steam generator tube leakageis assessed using Loss Threshold 4-.A.1.MNS Basis Reference(s):
- 1. NEI 99-01 CMT Integrity or Bypass Containment Loss 4.AIDocument No. Rev. 0 Page 257 of 272]
ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Containment Category:
D. CMT Integrity or BypassDegradation Threat: LossThreshold:
- 2. Indications of NCS leakage outside of containment Definition(s):
NoneBasis:Plant-Specific ECA-1.2 LOCA Outside Containment (ref. 1) provides instructions to identify and isolate aLOCA outside of the containment.
Potential NCS leak pathways outside containment include(ref. 1,2):* Residual Heat Removal (ND)" Safety Injection (NI)" Chemical
& Volume Control (NV)* NCP seals (NC)* PZR/NCS Loop sample lines (NM)GenericContainment sump, temperature, pressure and/or radiation levels will increase if reactorcoolant mass is leaking into the containment.
If these parameters have not increased, thenthe reactor coolant mass may be leaking outside of containment (i.e., a containment bypasssequence).
Increases in sump, temperature,
- pressure, flow and/or radiation level readingsoutside of the containment may indicate that the RGS-NCS mass is being lost outside ofcontainment.
Unexpected elevated readings and alarms on radiation monitors with detectors outsidecontainment should be corroborated with other available indications to confirm that the sourceis a loss of RGS-NCS mass outside of containment.
If the fuel clad barrier has not been lost,radiation monitor readings outside of containment may not increase significantly; however,other unexpected changes in sump levels, area temperatures or pressures, flow rates, etc.should be sufficient to determine if RG&-NCS mass is being lost outside of the containment.
Refer to the middle piping run of Figure O-F--41.
In this simplified
- example, a leak hasoccurred at a reducer on a pipe carrying reactor coolant in the Auxiliary Building.
Depending upon radiation monitor locations and sensitivities, the leakage could be detected by any of thefour monitors depicted in the figure and cause threshold 4-AD.1 to be met as well.Document No. Rev. 0 1 Page 258 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesTo ensure proper escalation of the emergency classification, the RGS-NCS leakage outside ofcontainment must be related to the mass loss that is causing the RGS-NCS Loss and/orPotential Loss threshold 47A.1 to be met.MNS Basis Reference(s):
- 1. EP/1 (2)/A/5000/ECA-1.2 LOCA Outside Containment
- 2. EP/1 (2)/A/5000/E-1 Loss of Reactor or Secondary Coolant3. NEI 99-01 CMT Integrity or Bypass Containment LossDocument No. Rev. 0 Page 259 of 272 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesFigure 1: Containment Integrity or Bypass ExamplesThreshold-Airbome--- --- ---release fromEffluent
',pathway.
MonitorVentRCPSealCoolinqI Document No. I Rev. 0 Page 260 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesContainment Barrier:Category:
D. CMT Integrity or BypassDegradation Threat: Potential LossThreshold:
- 11. Containment-RED Path conditions met IDefinition(s):
NoneBasis:Plant-Specific Critical Safety Function Status Tree (CSFST) Containment-RED path is entered if containment pressure is greater than or equal to 15 psig and represents an extreme challenge to safetyfunction.
(ref. 1).15 psig is based on the containment design pressure (ref. 2).GenericIf containment pressure exceeds the design pressure, there exists a potential to lose theContainment Barrier.
To reach this level, there must be an inadequate core cooling condition for an extended period of time; therefore, the RGS-NCS and Fuel Clad barriers would alreadybe lost. Thus, this threshold is a discriminator between a Site Area Emergency and GeneralEmergency since there is now a potential to lose the third barrier.MNS Basis Reference(s):
I1. EP/1(2)/A/5000/F-0 Critical Safety Function Status Trees2. UFSAR Section 6.2 Containment Systems3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.ADocument No. Rev. 0 Page 261 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesContainment Barrier:Category:
D. CMT Integrity or BypassDegradation Threat: Potential LossThreshold:
- 12. Containment hydrogen concentration
> 6%Definition(s):
NoneBasis:Plant-Specific Following a design basis accident, hydrogen gas may be generated inside the containment byreactions such as zirconium metal with water, corrosion of materials of construction andradiolysis of aqueous solution in the core and sump. (ref. 1).The lower limit of deflagration of hydrogen in air is > 6% and is the maximum concentration atwhich hydrogen igniters can be placed in service (ref. 2).To generate such levels of combustible gas, loss of the Fuel Clad and NCS barriers must haveoccurred.
With the Potential Loss of the containment
- barrier, the threshold hydrogenconcentration, therefore, will likely warrant declaration of a General Emergency.
GenericThe 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 lowerdeflagration limit). A hydrogen burn will raise containment pressure and could result incollateral equipment damage leading to a loss of containment integrity.
It therefore represents a potential loss of the Containment Barrier.MNS Basis Reference(s):
- 1. UFSAR Section 6.2 Containment Systems2. EP/I(2)/A/5000/FR-Z.4 Response to High Containment Hydrogen Concentration
- 3. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.BDocument No. I Rev. 0 Page 262 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Containment Category:
D. CMT Integrity or BypassDegradation Threat: Potential LossThreshold:
- 3. Containment pressure
> 3 psig with EITHER a failure of both trains of NS ORfailure of both trains of VX-CARF for -15 min. (Note 1)Note 1: The Emergency Coordinator/EOF Director should declare the event promptly upon determining that timelimit has been exceeded, or will likely be exceeded.
Definition(s):
NoneBasis:Plant-Specific The containment Phase B pressure setpoint (3 psig, ref. 1, 2) is the pressure at which thecontainment cooling systems should actuate and begin performing their function.
One full train of containment cooling operating per design is considered (ref. 1, 2):* One train of Containment Air Return Fan System (VX-CARF),
and" One train of Containment Spray System (NS)Once the Residual Heat Removal system is taking suction from the containment sump, withcontainment pressure greater than 3 psig and procedural
- guidance, one train of containment spray is manually aligned to the containment sump. If unable to place one NS train in serviceor without an operating train of VX-CARF (the CARF with a 10-minute delay) within 15 minutesa potential loss of containment exists. At this point a significant portion of the ice in the icecondenser would have melted and the NS system would be needed for containment pressurecontrol.
The potential loss of containment applies after automatic or manual alignment of thecontainment spray system has been attempted with containment pressure greater than 3 psigand less than one full train of NS is operating for greater than or equal to 15 minutes.The potential loss of containment also applies if containment pressure is greater than 3 psigand at least one train of VX-CARF is not operating after a 10 minute delay for greater than orequal to 15 minutes.
Without a single train of VX-CARF in service following actuation, thepotential loss should be credited regardless of whether ECCS is in injection or sumprecirculation mode after 15 minutes.GenericDocument No. Rev. 0 Page 263 of 272 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesThis threshold describes a condition where containment pressure is greater than the setpointat which containment energy (heat) removal systems are designed to automatically actuate,and less than one full train of equipment is capable of operating per design. The 15-minute criterion is included to allow operators time to manually start equipment that may not haveautomatically
- started, if possible.
This threshold represents a potential loss of containment inthat containment heat removal/depressurization systems (e.g., containment sprays, icecondenser fans, etc., but not including containment venting strategies) are either lost orperforming in a degraded manner.MNS Basis Reference(s):
- 1. MNS Technical Specification 3.6.62. MNS Technical Specification 3.6.6 Bases3. MNS Technical Specification 3.3.24. UFSAR Section 6.2 Containment Systems5. NEI 99-01 CMT Integrity or Bypass Containment Potential Loss 4.CI Document No. I Rev. 0 1 Page 264 of 2721 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Containment Category:
F. Emergency Coordinator JudgmentDegradation Threat: LossThreshold:
- 1. Any condition in the opinion of the Emergency Coordinator/EOF Director that indicates loss of the Containment barrierDefinition(s):
NoneBasis:Plant-Specific The Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the Primary Containment barrier is lost. Such a determination should includeimminent barrier degradation, barrier monitoring capability and dominant accident sequences.
" Imminent barrier degradation exists if the degradation will likely occur within two hoursbased on a projection of current safety system performance.
The term "imminent" refersto recognition of the inability to reach safety acceptance criteria before completion of allchecks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability
- concerns, readings fromportable instrumentation and consideration of offsite monitoring results." Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
GenericThis threshold addresses any other factors that may be used by the Emergency DeFeGtOtCoordinator in determining whether the Containment Barrier is lost.MNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment PC Loss 6.AI Document No. I Rev. 0 Page 265 of 272 1 ATTACHMENT 2Fission Product Barrier Loss/Potential Loss Matrix and BasesBarrier:
Containment Category:
F. Emergency Coordinator JudgmentDegradation Threat: Potential LossThreshold:
- 1. Any condition in the opinion of the Emergency Coordinator/EOF Director that indicates potential loss of the Containment barrierDefinition(s):
NoneBasis:Plant-Specific The Emergency Coordinator judgment threshold addresses any other factors relevant todetermining if the Primary Containment barrier is potentially lost. Such a determination shouldinclude imminent barrier degradation, barrier monitoring capability and dominant accidentsequences.
" Imminent barrier degradation exists if the degradation will likely occur within two hoursbased on a projection of current safety system performance.
The term "imminent" refersto recognition of the inability to reach safety acceptance criteria before completion of allchecks.* Barrier monitoring capability is decreased if there is a loss or lack of reliable indicators.
This assessment should include instrumentation operability
- concerns, readings fromportable instrumentation and consideration of offsite monitoring results." Dominant accident sequences lead to degradation of all fission product barriers andlikely entry to the EOPs. The Emergency Coordinator should be mindful of the Loss ofAC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.
GenericThis threshold addresses any other factors that may be used by the Emergency DieGteCoordinator in determining whether the Containment Barrier is lost.MNS Basis Reference(s):
- 1. NEI 99-01 Emergency Director Judgment PC Potential Loss 6.ADocument No. Rev. 0 Page 266 of 2721 ATTACHMENT 3Safe Operation
& Shutdown Areas Tables R-2 & H-2 BasesBackground NEI 99-01 Revision 6 ICs AA3 and HA5 prescribe declaration of an Alert based on impededaccess to rooms or areas (due to either area radiation levels or hazardous gas concentrations) where equipment necessary for normal plant operations, cooldown or shutdown is located.These areas are intended to be plant operating mode dependent.
Specifically the Developers Notes for AA3 and HA5 states:The "site-specific list of plant rooms or areas with entry-related mode applicability identified" should specify those rooms or areas that contain equipment which require a manual/local action as specified in operating procedures used for normal plant operation, cooldown andshutdown.
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, thelist should specify the plant mode(s) during which entry would be required for each room orarea.The list should not include rooms or areas for which entry is required solely to performactions of an administrative or record keeping nature (e.g., normal rounds or routineinspections).
- Further, as specified in IC HA5:The list need not include the Control Room if adequate engineered safety/design featuresare in place to preclude a Control Room evacuation due to the release of a hazardous gas.Such features may include, but are not limited to, capability to draw air from multiple airintakes at different and separate locations, inner and outer atmospheric boundaries, or thecapability to acquire and maintain positive pressure within the Control Room envelope.
I Document No. I Rev. 0 1 Page 267 of 272 ATTACHMENT 3Safe Operation
& Shutdown Areas Tables R-2 & H-2 BasesMNS Table R-2 and H-2 BasesA review of station operating procedures identified the following mode dependent in-plantactions and associated areas that are required for normal plant operation, cooldown orshutdown:
OP/1 &2/A/6100/003, Enclosure 4.2, Step3.8.4.1Perform OP/l/A/61 OO/SD 1(Prepare For Cooldown).
NIAN/ANoOP/1 &2/N6100/003, Perform NC System degas N/A N/A NoEnclosure 4.2, Step per OP/1/A/61OO/SD-10 3.8.6 (NC System, PRT andNCDT Degas).OP/I &2/A/6100/003, Open breakers on Transfer Yard 1 NoEnclosure 4.2, Steps Transformer Cooling3.8.8.1 & 3.8.9.1 Groups.OP/1&2/A/6100/003, Perform Main Steam Main Steam Doghouses 1 NoEnclosure 4.2, Step Safety Valve testing.3.8.14OP/1 &2/N6100/003, Check transfer of Aux Turbine Bldg. Basement I NoEnclosure 4.2, Step Steam from C htr Bleed to (739') North Wall3.8.17.1 Main Steam (Close 1SP-1(Main Steam to 1A CFPump Turb Isol) and 1SP-2(Main Steam to 1B CFPump Turb Isol).1AS-1 1).OP/I &2/N6100/003, Stop G HDT Pumps per Turbine Bldg. Basement 1 NoEnclosure 4.2, Step OP/Il/B/6250/004 (739') West Wall3.8.21 (Feedwater Heater Vents,Drains, and Bleed System).OP/I1 &2/A/6100/003, Stop C HDT Pumps per Turbine Bldg. Basement 1 NoEnclosure 4.2, Step OP/I/B/6250/004 (739') HP Heater Panel3.8.23 (Feedwater Heater Vents,Drains, and Bleed System).OP/1 &2/A/6100/003, Transfer of Aux Steam to Service Bldg. (739') or 1 NoEnclosure 4.2, Step Unit 2 or Aux Electric Auxiliary Boiler Room3.8.34 Boilers perOP/I/B/6250/007 B(Auxiliary Electric Boilers).
OP/I1 &2/A/6100/003, Close 1SP-1 (Main Steam Turbine Bldg. Mezz (760') 1 NoEnclosure 4.2, Step to 1A CF Pump Turb Isol) at CF Pumps3.12.7 and 1SP-2 (Main Steam to1B CF Pump Turb Isol).OP/1 &2/A/6100/003, Shutdown MG Sets per MG Set Room (767') 3 NoEnclosure 4.2, Step OP/1/A/6150/008 (Rod3.13.16.6 Control),
Enclosure 4.5IDocument No. I Rev. 0 F Page 268 of 272 1 ATTACHMENT 3Safe Operation
& Shutdown Areas Tables R-2 & H-2 BasesOP/1 &2/A/6100/003, Enclosure 4.2, Step3.14.4Secondary System WetLayup Chemical addition(Chemistry).
Secondary Chemistry Laband TB Basement (739')3NoI. IOP/1 &2/A/6100/003, Enclosure 4.2, Step3.14.6.2Begin performance ofstroke time testing of PzrPORVs.Inside Containment 3NoOP/1 &2/A/61 00/SD-1, Step 3.5When RP allows access toLower Containment, beginEnclosure 4.2 (Pre-Cooldown Containment Entry). This enclosure performs a Containment Inspection with RP,Engineering andOperations involvement.
Inside Containment 3NoOP/I &2/A/61 OO/SD- After required amount of Aux. Bldg. (NM Lab 716') 3 No1, Step 3.3.4 boron is added for SDM Counting Room (767')requirements for blockingP-11, Primary Chemistry samples NC System.OP/i &2/A/61 OO/SD- After required amount of Aux. Bldg. (NM Lab 716') 3 No1, Step 3.4.9 boron is added for SDM Counting Room (767')Shutdown BoronConcentration, PrimaryChemistry samples NCSystem.OP/I &2/N6100/SD-After required amount of Aux. Bldg. (NM Lab 716') 3 No1, Step 3.5.7 boron is added for Crud Counting Room (767')Burst Boron Concentration, Primary Chemistry samplesNC System.OP/I &2/A/61 00/SD- After required amount of Aux. Bldg. (NM Lab 716') 3 No1, Step 3.6.7 boron is added for Counting Room (767')Refueling BoronConcentration, PrimaryChemistry samples NCSystem.OP/I &2/A/61 00/SD- Have Radwaste align Aux. Bldg. (716') 1-3 No10, Step 3.5.1.1 Nitrogen for NCDT Degas Radwaste Areaper OP/I/A/6200/600 (WGSupport of Unit IShutdown).
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ATTACHMENT 3Safe Operation
& Shutdown Areas Tables R-2 & H-2 BasesOP/1 &2/N61 00/SD-10, Step 3.6.2Radwaste performs Phase1 PRT Degas perOP/OIA/62001518 (WasteGas Operation).
Aux. Bldg. (716')Radwaste Area1-3NoOP/1&2/AN6100/SD-Radwaste performs NC Aux. Bldg. (716') 1-3 No10, Step 3.7.1 System Degas per Radwaste AreaOP/1 &2/A/6200/600 (WGSupport Of Unit 1/2Shutdown).
OP/1&2/N6100/SD-Radwaste performs NCDT Aux. Bldg. (716') 1-3 No10, Step 3.8.3 Degas per Radwaste AreaOP/OIA/6200/518 (WasteGas Operation).
OP/I &2/A/61 00/SD- Radwaste performs Phase Aux. Bldg. (716') 1-3 No10, Step 3.9.1 & 2 PRT Degas per Radwaste Area3.9.2 OP/1 &2/N6200/600 (WGSupport Of Unit 1/2Shutdown) andOP/O/A16200/518 (WasteGas Operation).
OP/I &2/A/61 00/SD- Radwaste crossties BATs. Aux. Bldg. (733') BAT 3 No2, Step 3.2.3 AreaOP/i &2/A/61 OO/SD- When less than 1000 psig, Aux. Bldg. (733') Electrical 3 Yes2, Step 3.5 IAE places NC System Pene (733' and 750')Narrow Range PressureTransmitters in service.OP/I &2/A/6100/SD-To maximize charging flow, Aux. Bldg. (733') Ledge 3 No2, Enclosure 4.2, adjust NC Pump Seal Outside VCT RoomStep 3.2.3.2 Water Injection Throttles to8-10 gpm.OP/1&2/A/61OO/SD-Radwaste crossties BATs. Aux. Bldg. (733') BAT 3 No4, Step 3.2.3 AreaOP/I &2/A/61 00/SD- Perform plant shutdown ETA, ETB, Aux. Bldg. 3 Yes4, Steps 3.11.1, tagging.
(733' and 750') South End3.11.2, & 3.12.3OP/I &2/A/61 00/SD- To maximize charging flow, Aux. Bldg. (733') Ledge 3 No4, Enclosure 4.2, adjust NC Pump Seal Outside VCT RoomStep 3.10.2 Water Injection Throttles to8-10 gpm.OP/I &2/A/61 00/SD- Secondary Chemistry to Chemistry Lab and 4 No4, Step 3.15 check S/G sulfate meets Turbine Bldg. (739') Northchemistry criteria for Wallcontinued cooldown.
I Document No. Rev. 0 Page 270 of 272 1 ATTACHMENT 3Safe Operation
& Shutdown Areas Tables R-2 & H-2 BasesOP/1 &2/N61 00/SO-10, Step 3.8Stroke time testing ofPORVs perPT/I/A/4151/005 (NCValve Stroke Timing TestUsing Air). Enclosures 13.4, 13.5, 13.6Inside Containment 4NoOP/I&2/A/6100/SO-Rack out and tag one NV ETA (750'), ETB (733') 4 Yes10, Step 3.11.1, and both NI Pumps per3.11.2, 3.11.3 OP/O/N6350/008 (Operation of StationBreakers).
OP/I&2/A/6100/SO-Tag out PD Pump at Aux. Bldg. (750') North 4 Yes10, Step 3.11.6 1 MXK-F2C (Reciprocating EndCharging Pump No 1).OP/I&2/N6100/SO-If LTOP vent requirements Inside Containment 4 No10, Step 3.12 are to be satisfied bysecuring 1 NC-36B (PzrPORV) open, Maintenance gags 1 NC-36B (PzrPORV).OP/1&2/A/6100/SD-Unlock and close 1/2ND- P/C, RHole, near 1N1-185, 4 Yes6A(B), Step 3.3 119 (1/2 A ND ECCS Outside CAD 212 (716')Sump Suction Relief Inlet ABPC thru CAD Door,Isol #2). FF59 (716')OP/I &2/A/61 00/SD- Perform PT/l/A/4206/030 Aux Building Pipechase 4 No6A(B), Enclosure (Draining ECCS Sump (716')4.1, Step 3.6 Piping Drain Reservoir Train A), Enclosure 13.1(Draining ECCS SumpPiping Drain Reservoir Train A in Modes 1 4) andPT/1/A/4206/031 (Draining ECCS Sump Piping DrainReservoir Train B),Enclosure 13.1 (Draining ECCS Sump Piping DrainReservoir Train B in Modes1 4).OP/I &2/A/61 OO/SD- Monitor and shift NC Aux Building (716'/733')
at 4 No6A(B), Encl. 4..1, System Filters on high DP. NC Filters RoomStep 3.8OP/I &2/N61 00/SD- De-energize 1/2ND-68A (A ETA Aux. Bldg. (750') 4 Yes6A(B), Encl. 4.2 ND Pump & A Hx Miniflow)
Step 3.7 in the open position at1/2EMXA-F12B (1/2ND-68A).Document No.IRev. 0Page 271 of 272 1Document No. ReV. 0 Page 271 of 272 ATTACHMENT 3Safe Operation
& Shutdown Areas Tables R-2 & H-2 BasesOP/1 &2/A/61 00/SD-6A(B), Step 3.31.1Adjust flow through CationBed Demineralizer perOP/1/N6200/001 D(Chemical and VolumeControl SystemDemineralizers).
Aux Building (750') overDemin Pits4NoOP/1&2/NA6100/S0-De-energize 1/2ND-67B (B ETB Aux. Bldg. (733') 4 Yes6, Step 3.4.3 ND Pump & B Hx Miniflow) in the open position at1/2EMXBI-2C (1/2ND-67B).Table R-2 & H-2 ResultsTable R-2/1H-2 Safe Operation
& Shutdown RoomslAreas Bldg. Elevation Unit I RoomlArea Unit 2 RoomlArea ModesAuxiliary 716' P/C, RHole, near 1NI-185, ABPC thru CAD Door, FF59 4Outside CAD 212800 (1 EMXA) 820 (2EMXA) 3, 4Auxiliary 750'803 (1 ETA) 805 (2ETA) 3, 4702 (Elec. Pene.) 713 (Elec. Pene.) 3Auxiliary 733' 722 (1 EMXB-1) 724 (2EMXB-1) 3, 4705 (1ETB) 716 (2ETB) 3, 4Plant Operating Procedures Reviewed1.2.3.4.5.6.7.8.OP/1 &2/A/6100/003 OP/I&21A/6100/SD-1 OP/I &2/A/61 00/SD-2OP/I&21A/6100/SD-4 OP/I&2/A/6100/SD-10 OP/I &2/A/61 00/SD-6A(B)
OP/1 &2/A/61 00/SO-6OP/1 &2/A/61 00/SO-10I Document No. Rev. 0 Page 272 of 272