ML18026A629

Generation Station - Volume 13, Emergency Preparedness Procedures 13.1.1, Rev. 049; 13.1.1A, Rev. 033; 13.8.1, Rev. 038 and 13.14.11, Rev. 013

ML18026A629
Person / Time
Site:	Columbia
Issue date:	01/17/2018
From:	Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
13.1.1, Rev. 049, 13.1.1A, Rev. 033, 13.14.11, Rev. 013, 13.8.1, Rev. 038
Download: ML18026A629 (292)

Text

DISTRIBUTION -VOLUME 13- EMERGENCY PREPAREDNESS PROCEDURES Distribution.Date: 01/17/18 LMB Filed By:

Procedure Number Revision Procedure Number Revision Procedure Number Revision 13.1.1 049 13.1.1A 033 13.8.1 038 13.14.11 013 TO: ENERGY NORTHWEST EXTERNAL CONTROLLED COPY PROCEDURE HOLDERS The following documents have been revised and are to be inserted into your controlled copy manual and the superseded revisions removed and destroyed. No receipt acknowledgement is required for the listed document(s).

Should you have questions on this distribution please contact Kim Saenz, Records and Information Supervisor at (509)377-2492 or KDSaenz@energy-northwest.com.Error! Reference source not found.

Procedure Number Revision Procedure Number Revision Procedure Number Revision 13.1.1 049 13.1.1A 033 13.8.1 038 13.14.11 013 EXTERNAL DISTRIBUTION Copy# Location Procedures 26 ReQion IV - US Nuclear ReQulatorv Commission All 28 Region IV - US Nuclear Regulatory Commission All 52 State of Washington, Military Department All 55 Federal EmerQencv ManaQement AQencv (FEMA) All 57 Benton County - Department of Emergency Management All 75 Department of Health Radiation Protection All 87 Document Control Desk - NRC All 142 Hanford EOC/SMT All 164 OreQon State Department of EnerQV --~ All 223 Franklin County Emergency Management All 224 Washington State Department of Health - Office of Radiation Protection All J+iL/--'5 Updated 10/06/17 ti~~

DISTRIBUTION-VOLUME 13- EMERGENCY PREPAREDNESS PROCEDURES Distribution Date: 01/17/18 LMB Filed By:

Procedure Number Revision Procedure Number Revision Procedure Number Revision 13.1.1 049 13.1.1A 033 13.8.1 038 13.14.11 013 NOTE: PPM 13.1 .1 is printed in color, single sided, on 11" x 17" paper, folded, 3-hold punched, stapled on the top & distributed by Document Control.

NOTE: Distribution of the PPM 13.1.1 wall flow charts is controlled by PPM 13.1.1 A and distribution is performed by EP.

Updated 10/06/17

• RG1 .1 GENERAL EMERGENCY R..... of p.NOU& tadioedHtY r...a.r.g in ofllle dose pater hn 1,000 nwem TEDE or 5.000 rrn m lhyrold COE I 1 I 2 i 3 i

• i (1) Reading on JilX Table 3 effluent radiation monitor 5 I DEF) RS1.1 SITE AREA EMERGENCY R*lnM orPNOUI radtOaCfhlly'""""' in o"9lte doN .....,

l'lan 100 ,,nm TED£ or 500 mrem lhyroed COE I 1 J 2 i GT column "SAE" for GE 15 min .

3 J 4 (1) Reading on ~Table 3 effluent radiation monitor J 5 i DEF J I

RA1.1 I 1 I 2 ALERT

°'

R...... of gaseoua liquid radioadtYlty re*dllng in o....

dose greatM lhan 10 """'TEOE or 50 ITftm lhYfOld COE

) 3 ) * ) 5 (1) Reacing on ilrt Table 3 effluent radiation monitor GT column MALERT" for GE 15 min.

) DEF ) RU1 .1 R..._.

\

UNUSUAL EVENT 1t11t 00CM 1

o1,...... or lqYll radfOadMIW

~ to, I 2 60 rr--. or

) 3 I 4 lonfef (1) Reading on~ Table 3 effluent radiation monitor GT column MUEM for GE 60 min .

\ 5

....,flan I DEF I 2 llma*

GT colurm "GENERAL" for GE 15 min.

OR OR OR OR (2) Dose assessm ent using actual meteorology indica tes (2) Dose assessment using actual meteorology (2) Dose assessment using actual meteorology indicates (2) Sample analyses for a gaseous or liquid release doses GT 1,000 1TWem TEOE or GT 5000 mrem thyroid indicates doses GT 100 mrem TEDE or GT 500 doses GT 10 nYem TEDE or GT 50 nYem thyroid COE indicates a concentration or release rate > 2 x ODCM COE at or beyond the SITE BOUNDARY rnrem thyroid COE at or beyond the SITE al or beyond the SITE BOUNDARY limits for GE 60 min.

(Notes 1, 2, 3, 4) BOUNDARY (Notes 1, 2, 3, 4) (Notes 1, 2, 3)

(Notes 1, 2, 3, 4)

I I I I I DEF I 1 RG1 .2 I 1 j 2 I 3 ) 4 / 5 I DEF I RS1 .2 I 1 I 2 j 3 j 4 j 5 j DEF I RA1 .2 j 1 2 3 4 5 Analysis of a liquid effluent sample indicates a concen tration Rad Field survey results Indicate EITHER of the following at or Field survey results indcate EITHER of the following at or or release rate that would r esult in doses GT 10 mrem TEDE beyond the SITE BOUNDARY: beyond the SITE BOUNDARY : or GT 50 mrem thyroid COE at or beyond the SITE

• Closed window dose rates GT 100 mR/hr expected BOUNDARY fDf 60 min. of exposure (Notes 1, 2) e Closed 'Nindow dose rates GT 1,000 mR/hr expected to continue for GE 60 min . to continue for GE 60 min.

e Analyses of fiekf survey sall'1)1es indicate thyrokt e Analyses of field survey samples indicate thyroid RA1 .3 I 1 I 2 I 3 I 4 ) 5 I DEF I COE GT 5,000 mrem for 60 min . of inhalation . COE GT 500 mfem for 60 min. of inhalation .

Field survey results indicate EITHER of the follo'Ning at or (Notes1 , 2) (Notes 1, 2) beyond the SITE BOUNDARY:

Closed window dose rates GT 10 mRhlr expected to continu e for GE 60 min .

e Analyses of field survey samples indicate thyroid CDE GT 50 m--em for 60 min . of inhalation .

A-RRad Spent fuel pool leve4 cannot be rettored lo at lent lh* top of the 1\191 raCQ fur 60 min..ies o, long<<

Sperw klel pool le..,el al the 109 of the Ml rKks (Notes 1, 2)

Signif'lcant k:J#erlng of water level aboWt, or da!Nge to, irrad*tedfuel Lovels I RG2.1 J 1 J 2 i 3 I

• J 5 [ DEF J RS2.1 J 1 J 2 J 3 J 4 J 5 J DEF J RA2.1 ) 1 i 2 ) 3 ) 4 J 5 I DEF J RU2 .1 I 1 J 2 i 3 J

• i 5 J DEF J Rad Spent fuel pool level cannot be restored to at least 0.5 ft Lowering of spent fuel pool level to 0.5 fl Uncovery of irradiated fuel in the REFUELING PATHWAY UNPLANNED water level drop in the REFUELING PATHWAY for GE 60 min. (Note 1) as indicated by EITHER of the following:

RA2.2 I 1 j 2 I 3 I 4 I 5 I DEF I

• SFP level LE 22 .3 ft 2 Effluent Monitor Cl.Hsification Thresholds Damage to in adiated fuel resulting in a release of radioactivity AND SFP low level alarm AND UNPLANNED rise in area radi ation levels as indicated by~

Fuel E11ent Reiease Point Monitor General SAE Alert UE High alarm on !!r£ of the following radiatk>n monitors: o f the following radiation monitors:

PRM-RE-11 ARM-RIS-1 Reactor Building Fuel Pool fvea ARM-RJS-1 Reactor Buiding Fuel Pool Area 3.05E-031,1CVcc

. React.a, Buidng Ex~uat PRM*RE-12 PRM-RE-13 7.50E<-02µCilec 7.50E*1 µCil'cc 2.82E*1 µCi/cc ARM-RI S-2 Reactor Building Fu el Pool fvea

• ARM-RIS-34 Reactor Building Elevation 606

• REA-RIS-609A-0 Rx Bldg Vent ARM-RIS-2 Reactor Bu~ding Fuel Pool Area ARM-RIS-J.4 Reactor Buiding Elevation 606 I

Cl Turbine Building Exhll~t RactwuteBl.llding E*h1u1t TEA*RIS-13 WEA-RIS-14 8.35E-02µCi/cc 3.45E-01 l,ICl/cc 8.35E-031,1CUcc 345E-021,1CUcc 8.35E..CW1,1Clfcc 3.45E-031,1Cllcc 4.22E-051,1CVcc 3_98E..CWl,ICUtc RA2.3 I 1 I 2 I 3 Lowering of spent fuel pool level to 10 ft I 4 I 5 I DEF I

[

R1dwu1, Eflk>>nl FDR-RIS~ 2XHl*Hlalarm Radilllion !eve.. that IMPEDE accns IO equc,merw n<<euary fur normal plant opwabOM, eooldown o, .tudown

] TSWEflu,111 TSW-RIS-5 3.00E-051,1Cl/cc 3 3' Servic,Wat,rProces1A SW-RIS-604 1.00E+02 Cpl RA3.1 I 1 i 2 ) 3 ) 4 ) 5 ) DEF)

Ana S1rviceWaterProcN1B SW-RIS-005 1.00E+02 Cpl (1) Dose rates GT 15 mRnir in Control Room Radiation (ARM-RI S- 19) or CAS (by survey)

L..ets OR (2) An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to .mx Table 9 Table 9 Safe Operation & Shutdown Rooms/Arus rooms or areas (Note 5)

Oam~e lo a lNded CHk CONFINEMENT BOUNDARY Room/fvea Modes Applicability J eu1 .1 ~I___ s,_o_rag~*-O~pe_rat_oo_,_ _ ~

R~467' Radwute Control Room (RHR flush W RW ta~~ _

Damage to a loaded canister (MPC) CONFINEMENT E 1 RW 467' Vrtal Island (RHR-V-9 disconnect)

RB 422' B RHR Pump Rm (local pump temperalllrn)

BOUNDARY as indicated by measured dose rates on a loaded overpaclt GT EITHER:

Conflnem11nt ISFSI RB 4S4' B RHR Pump Rm (operate RHR-V~5B) 20 nYemlhr (gamma + neutron) on the top of the Boundary overpack 100 nYemlhr (gamma + neutron) on the side of the overpack, excluding inlet and outlet Wets HOSTILE ACTION wilhin the PROTECTED AREA HOSTILE ACTION within the OINNER CONTROlLEO AREA or Co111lrmed SECURITY CONDITION or u...at eilbome attack thre,t within 30 mirute, HS1 .1 I 1 ) 2 I 3 I 4 I 5 I DEF I HA1 .1 I 1 I 2 I 3 I 4 I 5 I DEF I HU1.1 J 1 ) 2 i 3 J

• I 5 J DEF J A HOSTILE ACTI ON is ocwrring or has occurred 'Mthin (1) A HOSTILE ACTION is occurring or has occurred (1) A SECURITY CONDITION that does !!2!involve a the PROTECTED AREA as reported by the Security within the 0\1\NER CONTROLLED AREA as reported HOSTILE ACTION as reported by the Security 1 Sergeant or Security Lieutenant by the Security Sergeant or Securi ty Lieutenant OR Sergeant or Security Lieutenant OR Security (2) A validated notification from NRC o f an aircraft attack (2) Notification of a aecible security threat directed al the threat 'Mthin 30 min. of the site site OR (3) A validated notification tom the NRC providing information of an aircraft threat Seismic *"'*"' GT OBE levei.

2 rse-;'cA6~MA8.1 for po.;;ti alfor - 1 HU2.1 j 1 j 2 I 3 I 4 I 5 \ DEF I Seismic 1~:-;~;~e::r1:b~=~~e:rd~9:~:d I Seisrric event GT Operating Basis Earthquake (DBE) as Event L--------- Indicated by H13 .P851 .S l .5-1 (OPERATING BASIS EARTHQUAKE EXCEEDED) activated Notes Hazardouae... ent 1 The Emergency Director should declare the event HU3 .1 j 1 I 2 j 3 I 4 j 5 j DEF j promptly upon determining that time limit has been (1) A tornado strike within the PROTECTED AREA exceeded , or 'MIi iikeiy be exceeded OR 2 If an ongoing release is detected and the release start (2) Volcanic ash fallout requiring plant shutdown time is unknown , assume that the release duration has exceeded the specified time Hmlt 3 If the efflu ent flow past an effluent monitor is knO'M'l to HU3.2 I 1 I 2 I 3 I 4 \ 5 I DEF I 1

3 have stopped, indicating that the release p ath Is isolated, the effluent monitor reading ts no longer VALID for r Se~A6 .1/MA8.1 forpotenti~ -

I~:;~~~e: :~'!r~=~~e°:r:O:a~:d I Internal room or area FLOODING of a mag-iitude sufficient to require manual or automatic electrical isolation of a SAFETY Natural or classi fi cation purposes SYSTEM component needed for the current operating mode Tech . 4 The pre-calculated effluent monitor values presented in L---------

Hazard EAl..s RA1 .1, RS1 .1 and RGl .1 should be used for HU3 .3 J 1 J 2 i 3 J

• I 5 J DEF J emergency dassification assessments until the results (1) Movement of personnel 'Mthin the PROTECTED AREA is tom a dose assessment using actual meteorology are IMPEDED due to an offslte event involving hazardous available materials (e.g., an offsite chemical spill , 618-11 event or 5 If the equipment in the listed room or are a was already toxic gas release) inoperable or out-of-service before the event occurred ,

OR then no emergency classification is warranted (2) A h azardous event that r esults in on-site conditions 6 If CO NTAI NMENT CLOSURE is re-established prior to sufficient to prohibit the plant staff from accessing the site exceeding the JO-minute time limit, declaration of a via personal vehicles (Note 7)

General Emer gency Is not required 7 This EAL does not appty to routine traffic impediments such as fog, snow, Ice , or vehicle breakdO'M'ls or Table 5 accidents Pfant Structures Containing Safe Shutdown System s or 8 A manual scram action is any operator action . or set of Components HU4.1 I 1 ) 2 ( 3 ) 4 ) 5 I DEF )

actions, which causes the control ro ds to be rapidly A FIRE is [l2t extinguished 'Mthin 15 rrin. of I.DX of the inserted into the core, and does not include manually Vital portions of the Rad Was111/Co11trol Building following FIRE detection indications (Note 1):

driving in control rods or if11)1ementation of boron injection

• Report from the field (i.e., visual observation) 467' elevallon vrtal island strategies

• Receipt of multiple (more than 1) fire alarms or 487' elevalloo cable 'J)tHding room Indications H 9 If the affected SAFETY SYSTEM train was already inoperable or out of seMce before the hazardous event ocwrred , then emergency classification is not warranted Main Control Room and ver11cal cable chase 525' elevallon HVAC UH

• Field veriication of a single fire alarm AND Hazards The FIRE is located within filri Table 5 area 10 If the hazardous event only resulted in VISIBLE Reactor Building DAMAGE. wtth no indications of degraded performance Vital portions of the Turbine Building HU4.2 I 1 I 2 I 3 ( 4 [ 5 I PEE I 4 to at least one train of a SAFETY SYSTEM , then this emergen cy dassification is not warranted DEH prenure swrtches RPS ,wrtches on turbine throttle valves Receipt of a single fire alarm (i.e., Il2 other indications of a FIRE)

Fire AND Main steam line radiation monitors The fire alarm is indicating a FIRE within ADY. Table 5 area Turbine Building ventilation radiation monitora AND The existence of a FIRE is n ot verified within 30 min . of alarm Main ,teem tine piping up to MS-V-146 and the first stop valve, receipt (Note 1) -

Standby Serv,oe Watar Pump Houses Oieae1 Generator Building HU4.3 j 1 j 2 j 3 j 4 I 5 I DEF I (1) A FIRE wtthin the ISFSI or plant PROTECTEDAREA[l2t extinguished within 60 min. of the initial report, alarm or r se~A6-:VMA8.1 for po-;.:;-tiifor - 1 indication (Note 1)

OR Tab~ 9 Safe Operation & Shutdown Rooms/Areas I ~::~~~~ :!~ri:~:~e':d:ag:~:d I (2) A FIRE wi!hjn the ISFSI °' plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Room/Alea Modes Applicability GaHou. releaM IMPEDING acceu to eqlapment naceaury lot 110,mal plllnl operations, cooldown Of ,huldown

~467' Radwute Control Room (RHR flU$h to R ~) _ _ 3 5 RW 467' Vital Island (RHR-V-9 disconnect) 3 HA S.1 [ 1 J 2 I 3 I 4 I 5 Release of a toxic, corrosive , asphyxiant or flammable gas

[ DEF I Hazardous ~ RHR Pump Rm .(local pumptempe~ ___l Gases into .e Table 9 rooms or areas RB 454' B RHR Pump Rm (operate RHR-V-85B) 3 AND Entry into the room or area is prohibited or IMPEDED (Note 5)

HS6.1 lnabltity lo control

• key aafety function ff'om ou19lde the Control I 1 I 2 j 3 I 4 I 5 iW9 HA6 .1 Control Room evacuation resulting 111 trao,fer of plant control to altem1teloc1tion, I 1 I 2 j 3 I 4 I 5 I DEF I 6 An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel or Alternate Remote Shutdown Panel An event has resulted in plant control being transferred tom the Control Room to the Remote Shutdown Panel or

~emate Remote Shutdown Panel Control AND Room Control of !!It of the following key safety functions Is n2!

Evacuation re established 'Mthin 15 min. (Note 1):

• Reactivity (Modes 1 and 2 onty)

• RPV water level

• RCS heat r emoval ache, condrtion1 ellilti111 wl'ich In the judgment of the Eme,vency Other oondtlioM H itting whk:h In the Ju:lgmerw of the Emergency O!her condibo"' exi911,v which In the ~ment of the Other conditions Hi.ting wtwctl In the Judgment ol the Otrector wa1ra11t declarabOn of General Emergency Dmtclor warrant dee.lat Ilion of Sile Area Emergency Emergency Director warrant declaration of 111 Alert Emergency Oirectorwa,rant deciaiation or a UE HG7.1 J 1 J 2 i 3 J

• J 5 J DEF J HS7.1 I 1 J 2 J 3 J 4 J 5 J DEF J HA7.1 I 1 J 2 J 3 J 4 J 5 J DEF J HU7 .1 I 1 I 2 I 3 J 4 I 5 I DEF I Other conditions exist which , in the judgment of the Other conditions exist which, in the judgment of the Other conditions exist which , in the judgment of the Other conditions exist 'Mlich, in the judgment of the Emergency Director , indicate that events are in progress or Emergency Director, lncicate that events are in progress or Emergency Director, indicate that events are in progress or Emergency Director, indicate that events are ln progress or have occurred which involve actual or IMMINENT have occurred which invONe actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the 7 substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk level of safety of the plant or indicate a security threat to fa cility protection has been initiated. No releases of Judgment results In an actual loss of physical control of the facility. malicious acts, (1) toward site personnel or equipment that to site personnel or damage to site equipment because of radioactive material requiring offsite response or monitoring Releases can b e rea sonably expected to exceed EPA could lead to the likely faihxe of or, (2) that prevent effective HOSTILE ACTION. Any releases are expected to be limited are expected unless further degradation of SAFETY Protective Action Guideline exposure levels offslte for more access to equipment needed for the protection of the public. to small fr actions of the EPA Protective Action Guideline SYSTEMS occurs.

than the immediate site area. Ally releases are not expected to result ln exposure levels exposure levels .

which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.

13.1.1 Rev. 49 CLASSIFYING THE EMERGENCY Modes: .____1___.I ENERGY l..__2___.I _I_3~I _I_4~I _I_5~I ! DEF NORTHWEST 1/17/2018 Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled ALL CONDITIONS

• CG1 .1 GENERAL EMERGENCY I

• I s I RPV level LT - 161 in. for GE 30 min. (Note 1)

AND SITE AREA EMERvcNCY CS1.1 (1) CONTAINMENT CLOSURE AND I 4 I s I ill?! established CA1 .1 ALERT I

• I s (1) Loss .of RPV inventory as indicated by RPV level LT -SO in CU1.1 UNUSUAL EVENT I

• I s

{1) UNPLANNED loss of reader coolant resuns u, RPV level less than a required lower limit for GE 15 min (Note 1) 6rJ::J. of the following indications of containment challenge: OR OR RPV level LT -129 in.

(2)RPV level~ be monitored for GE 15 min. (Note 1) (2) RPV level~ be monitored CONTAINMENT CLOSURE D21 established {Note 6) OR AND AND Explosive mixture inside PC (2) CONTAINMENT CLOSURE established UNPLANNED increase in i.QY Table 1 sump or pool levels (H2 GE 6% and 02 GE 5%)

AND UNPLANNED ina-ease in w. Table 1 sump or pool due to a loss of RPV inventOJY UNPLANNED rise in PC pressure levels due to a loss of RPV inventory RPV level LT -161 in.

RB area radiation GT iDY Maximum Safe Operating level (PPM 5.3.1 Table 24)

CS1.2 I

• I s I CG1.2 I 4 I s I RPV level-'i!.!l!lQ1 be monitored for GE 30 min (Note 1) 1 RPV level ann21 be monitored for GE 30 min. (Note 1)

AND AND Table 1 SumpalPool RPV Core uncovery Is indicated by .fillX of the foll owing:

Lllnl Core uncovery is Indicated by i.!ll'.. of the following:

UNPLANNED wetwell level rise GT 2 inches UNPLANNED wetwell level rise GT 2 inches (PPM 5 2 1 entry condition) Air£ valid Hi-Hi level alarm on R* 1 (PPM 5 2. 1 entry condition) VALID indication of RB room nood1ng as Identified by through R-5 sumps VALID Indication of RB room flooding as identified by high level alarms (PPM 5 3 1 Table 25) EDR GE 25 GPM Offslte high level alarms (PPM 5 3.1 Table 25) Ob6erva11on of UNISOLABLE RCS leakage outside Observation of UNISOLABLE RCS leakage outside FDR GE 10GPM Startup Transformer TR-S pnmary containment of sufficient magnitude to indicate primary containment of sutr1eient magnitude to inchcate core uncovery Wetwell level rise Backup Transformer TR-B core uncovery Observation of UNISOLABLE RCS Backfeed 500 KV power through M111r1 AND leakage Transformers (rf already aligned in

&u of the following indications of containment challenge modes 4 , 5 , def only)

CONTAINMENT CLOSUREoQt. established (Note 6)

Explosive mixture inside PC Onslte (H2 GE 6% and 0 2 GE 5%) DG1 UNPLANNED rise in PC pressure DG2 RB area radiation GT iill Maximum Safe Operating Main Generator via TR-N1/N2 level (PPM 5.3.1 Table 24 )

2 C A2.1 LOMofll.olflllllndtlaneleACpowll"lo . . . . . . . lMala for ISIIWMMor-.,

I 4 j 5 I DEF I Loss of ill ortsite and ill onsite AC power capab1hty to CU2 .1 lANof.WOMA.CpoMl'NUlet .. .,,.,...,, ..... fDr15 l4I S 1 DEF I AC power capability, Table 2, to emergency buses SM-7 Lou d emergency buses SM-7 and SM-8 f0< GE 15 min. (Note 1) and SM-8 reduced to a single pov.oer source for GE 15 min.

C l"-1 AC'-

(Note 1)

~

AND additional single power source failure will result in a loss COld SIII of ill AC JX)Wer to SAFETY SYSTEMS

~

I Table 7 RCS Reheat Duration Threeholcla CAJ .1 I 4 I s CUJ .1 I 4 I s 3 II 1n RCS hell removal system Is In operation wilhln this time frame and RCS temperature Is being reduced lhe EAL UNPLANNED increase in RCS temperature to GT 200°F for GT Table 7 duration (Note 1)

UNPLANNED increase in RCS temperature to GT 20o*F RCS Temp. 11n21.1pplicable OR Contai nment Heat-up UNPLANNED RPV pressure Increase GT 10 psig CUJ.2 I 4 I s I RCS Status Closure Status Duration Loss o f ill RCS temperature and RPV water level tndicatton f0< GE 15 min. (Note 1)

Intact NIA GO min*

Loa ol Ytal DC power ror 15 "*--- or longer 4

establrshed 20min

• Loud

.t::l2J.lntad CU4.1 I 4 I s I D21 established Omin Indicated voltage LT 108 voe o n ~ 125 voe buses VltalDC DP-S1-1 and DP-51-2 for GE 15 min (Note 1)

CUS.1 j 4 I 5 I DEF j 5 System Onslte ORO NRC Loss of ill Table 4 onsite communication methods OR Plant Public Address (PA) System X Loss of ill Table 4 ORO communication methods OR Plant Telephone System X X Loss of ill Table 4 NRC communication methods Plant Radio System Operations and X Hazantoua *Mnl llltdlng a SAFE TY SYSTEM nMdtd tor ....

<UTMt operatq modi Security Channels CA 6.1 I 4 I s I Oftsite caning capability from the X X The occurrence of .i!lY. Table 8 hazardous event Control Room via direct telephone AND 6 Long distance calling capabllily on

!he commercial phone system X X Seismic event Internal Of external FLOODING event Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode High winds AND EITHER Event damage has caused indications of degraded Tornado strike perf0<mance to a second train of a SAFETY SYSTE M FIRE needed for !he current operating mode EXPLOSION OR Event damage has resulted in VISIBLE DAMAGE to a Volcanic ash fallout second train of a SAFETY SYSTEM needed for the Other events with similar hazard current operating mode charactenstics as determined b)' the Shift (Notes 9. 10)

Manager 13.1.1 Rev. 49 CLASSIFYING TiiE EMERGENCY Modes: .___ 4 __.I I 5 11 DEF I ENERGY 1117/2018 Cold Shutdown Refueling Defueled NORTHWEST COLD CONDITIONS (RCS :S 200°F)

• r GENERAL EMERGENCY MG1 .1 I 1 I 2 I 3 j

- ~ 1"* *.. ,1~~;1,.*"w,*-..,..i:~"..,. "'/~1 *~.....

Prolonged Jou o l lJ1 offaite and Ill onsile AC power 10 emet"gency bl.$el Loss of Al! offsite AND fill onsite AC power capability to I

MS1 .1

~~ff<<*~~

SITE AREA EMERGENCY Lon ol all offaite and .Ill onsite AC po-r lo eme,gency buses for 15minuteaorlonger j 1 I 2 I 3 I Loss of fill offsite an d fil! onsite AC power capability to I

MA1.1 ALERT

~I~1~1~2~1~,~i-~-~-~

AC power capability, Table 2 , to emergency buses SM-7

, , , ~ ~ }. . Jlq.r.-ott,._,.,~\*~1,,._.e~*

Loss of Iii bul o ne AC power 1oun::e 10 emerge ncy- bu.es lor15minutesorlonger _ ,

I MU1.1 I UNUSUAL EVENT 1 I 2 I 3 I

"""~~""'""'

loM of II offlife AC power eaip abll~ to emergency- bl.II" for 1!5minutes orlonger Loss of Al! offsite AC power capability , Table 2, to emergency

• • ~°"' '!I emergency buses SM-7 and SM-8 emergency buses SM -7 and SM-8 for GE 15 min. (Note 1) and SM-8 reduced to a single power source for GE 15 min . buses SM-7 and SM-8 for GE 15 min . (Note 1)

AND EITHER: (Note 1) 1 Restoration of emergenc.y bus SM-7 or SM-8 in LT 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is n.2! likely (Note 1)

OR AND

~ additional single power source failure 'MIi result in a l oss of Al! AC power to SAFETY SYSTEMS Table 2 AC Power Sources Loss of Offsite Emergenei RPV level cannot be restored and maintained AC Power GT -186 in .

i-- -Lo"""~,'"',""11,...,-

- ,....,,.""",- A

...,c-,-""...,""~'...,'D

...,C_po_w-,,-.,-~-'"~ .,~,15 . Startup Transformer TR- S Backup Transformer TR-B MG1.2 minute101longer

• - - Backfeed 500 KV power through Main Transformer s (if already aligned in modes 4, 5, def only)

Loss of~ offsite AND fill onsite AC power capability to Onsite emergen cy buses SM-7 and SM-8 for GE 15 min. (Note 1)

DG1 AND DG2 Indicated voltage is LT 108 voe on l22ttl 125 voe buses Lon ol Jll vilal DC pow~ for 15 min~es or longer 2 DP-S1 -1 and DP-S 1-2 for GE 15 min . (Note 1)

MS2.1 I 1 I 2 I 3 i Main Generator via TR-N1/N2 Loss of Vital DC Indicated voltage is LT 108 voe on both 125 voe buses Power DP-S1 -1 and DP- S1-2 for GE 15 min . (Note 1) 3 I= UNPLANNED loa of Control Roo m indicalions for 15 minutes or tonge r W11ha significanltransien!Slprogren rl~1,.::i; ::;2;::::l; ::;3:::;i=::;=:::;:=~- -t":M AA UNPLANNED event results in the inability to monitor one

U UNPlANNED loss ofConlr ot Room Indications for 15 minutes or longer

3:":.1:--;::

I : 1;:::::l;::::;2;:::::l:::;3::;l;::::::::;:;:::::::;:;:::::~ - "4 M UNPLAN NED event results in the inability to monitor or more Table 10 parameters from 'Mthin the Conb'ot Room on e or more Table 10 parameters from within the Control Loss of for GE 15 min. (Note 1) Room for GE 15 min . (Note 1)

Control AND Room  !!::!r:J. Table 11 transient event in progress Indications Table 10 Safety System Pa ramete rs Rnctor coolant actf'lify greater than Technical Speclficallon Reactor power allowablellmils Table 5 Plant Sbuctures Containing Sare Shutdown Systems or RPVlevel Com~nents - -*- RPV pressure MU4.1 I 1 j 2 I 3 I Primary containm ent pressure SJAE CONDSR OUTLET RAD HI -HI alarm (P602) 4 Vital portions of the Rad \Naste /Control Building:

467' elevation vital island Wetwell level Wetwell temperature RC S 487' eleva tion cable spreading room AcU vlty Main Control Room and vertica l cable chase MU4.2 I 1 I 2 I 3 j 525' elevation HVAC area Table 11 Trans ient Events Coolant activity GT 0.2 µCi/gm dose equivalent 1-131 Reactor Building Vi tal portions of the Turbine Building Reactor scram M

Systam Matfunct.

OEH pressure switches RPS swilches on 11.Jrbine throttle valves Main steam line radia tion monitors Runback GT 25% thermal read:or power Electrical load rej ection GT 25% full electrical load MU5.1 RCS leakage for 15 min~e* or longer I 1 ) 2 I 3

( 1) RCS unidentified or pressure boundary leakage I -

5 Turbine Building ventilation radiation monitors Mai n steam line piping up to MS-V-1 46 and the firs! stop valves ECCS injection Thermal power oscillations GT 10%

GT 10 gpm for GE 15 min .

OR RCS Leakage Standby- Service \Nater Pump Houses (2) RCS identified leakage GT 25 gpm for GE 15 min .

Diesel Generator Building OR (3) Leakage from the RCS to a location outside containment GT 25 gpm for GE 15 min .

lnability101hutdownthereac1orcau.iooacha*engeto RP\I Automaiic or mar..ial scram falls lo shut d own the reaclor , and Automatic or manual scram fails to shut down the reactor

,_ water level or RCS heal removal

--1--

subsequent manual actions taken at the reactor control consoles are.ams1JCCessfulin1twtllngdown l he reactor MS6.1 I 1 I 2 I MA6.1 I 1 I 2 i MU6 .1 j 1 I 2 M automatic OR manual scram fails to shut down the An automatic OR manual scram fail s to shut down the AA automatic OR manual scram did !!2! shut down the reactor reactor reactor 6 AND 8!! actions to shut do'Ml the reactor are nQ! successful as AND Manual saam actions taken at the reactor control console AND A subsequent automatic saam OR manual scram action RPS indicated by reactor power GT 5%

Failure (mode switch in shutdo'Ml , manual push buttons or ARI) are taken at the reactor control console (mode switch in

!lfil successful in shutting down the reactor as indicated by shutdown, manual push buttons or ARI) is successful in

- - - - - - - - - - . - - - - - -..1..-. AN~!~Tl:~e~:cannot be restor ed and maintained reactor power GT 5% (Note 8) shutting down the reactor as indicated by reactor power LE Table 4 Comroonication Methods above -186in.Of cannot be determined 5% (APRM downscale) (Note 8)

OR System Onsite ORO NRC W,N temperature and RPV pressure~ be maintain ed below the HCTL

' -- --I-_. Plant Public Address (PA) System X

~ 1111.onsiteoro"litecommunicahonscapab ilitles Plant Telephone System X X MU7.1 I 1 I 2 j 3 j 7 Plant Radio System Operations and Security Channels X (1) Loss of Al! Table 4 onsite communication methods OR Loss of (2) Loss of .fill Table 4 ORO communication methods Comm. Offsite calling cap ability from the X X Control Room via direct telephone OR (3) Loss of !ft Table 4 NRC communication methods

' - - - -,1 Long distance calling capability on X X the commercial phone system Haza,dous event affecting a SAFETY SYSTEM needed forlhe I- CUll~ ngm..=:....._

MAB.1 I 1 I 2 I 3 I The occurrence of aw. Table 8 hazardous event Table 8 Hazardous Events AND Event damage has caused in dications of degraded 8 Seismic event Internal or extern al FLOODING event performance on one train of a SAFETY SYSTEM needed for the current operating mode Hazardous AND EITHER :

Event Hig-1 winds Event damage has caused indications of degraded Affecting Tornado strike performance to a second train of a SAFETY SYSTEM Safety FIRE needed for the curr ent operating mode Systems OR EXPLOSION Volcanic ash fallout Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the Other events Ylith similar h azard current operating mode ch ara cteristics as determined by the Shift (Notes 9, 10)

Manager FG1.1 I 1 j 2 I 3 FS1 .1 I 1 I 2 j 3 FA1.1 I 1 I 2 I 3 F Loss of JllX two barriers Loss or potential loss of J.DX two barriers (Table F-1) 8w loss or Jm'. potential loss of EITHER Fuel Clad or RCS Fission Product AND barrier (Table F-1)

Barrier Degradation Loss or potential loss of the third barrier (Table F- 1)

Table F-1 Fission Product Barrier Threshold Matrix FC - Fuel Clad Barrier RCS - ~eactor Coolant System Barrier PC - Containment Barrier Loss Potential Loss Loss Potential Loss Loss Potential Loss RPV level QilD.Il.Ql be restored and RPV level cannot be restored and A SAG entry required maintained GT -161 in . maintained GT -161 in . SAG entry required RPV Water Level or cannot be determined. or cannot be determined .

UNI SOLABLE break in fillY. of the UNISOLABLE pri mary syste m leakage UNI SOLABLE primary system leakage follo'Mng:

• Main Steam Line tha t results in exceeding EITHER: that results in exceeding EITHER:

• RCIC Steam Line RB area temperature alarm level (PPM RB area maximum safe operating B

• RWCU 5.3 .1 Table 23) temperature (PPM 5.3.1 Table 23)

RCS L eak Rate

• Feedwater OR OR OR RB area radiation alarm level (PPM RB area maximum safe operating Emergency RPV Depressurization is 5.3. 1 Table 24) radiation (PPM 5.3 .1 Table 24) required PC pressure GT 45 psig UNPLANNED rapid drop in PC pressure OR fo llowing PC pressure rise Explosive mixture exists inside PC C PC pressure GT 1.68 psig due to RCS (H 2 GE 6% and 0 2 GE 5%)

OR PC Conditions leakage OR PC pressure response n21, consistent with LOCA conditions VWV temperature and RPV pressure cannot be maintained below the HCTL Containment Radiation Monitor CMS-R1S-27E or CMS-RIS-27F D reading GT 3,600 RA,r Containment Radiation Monitor Contai nment R adiation Monitor CMS-RIS-27E or CMS-RIS-27F CMS-R1S-27 E or CMS-R1S-27F PC Rad / OR reading GT 70 R/hr reading GT 14,000 R/hr RCS Activity Primary coolant activity GT 300 µCi/gm Dose Equivalent 1-131 UNI SOLABLE direct do'N!lstream E pathway to the environment exists after PC Integrity or PC isolation signal Bypass OR Intentional PC venting per EOPs F 8nY. condition in the opinion of the f::JJ:i.. condi tion in the opinion of the 8nY. condition in the opinion of the 8nY. condition in the opinion of the 8aY condition in the opinion of the 8nY. condition in the opinion of the Emergency Emergency Director that indicates loss , Emergency Director that indicates Emergency Director that indica tes Emerg ency Director that indicates loss Emergency Director that indicates Emergency Director that indicates loss potential loss of the Containment Director of the fuel clad b arrier potential loss of the Fuel Clad barrier of the RCS barrier j potential loss of the RCS barrier of the Containment barrier Judgment barrier 13.1.1 Rev. 49 CLASSIFYING TH E EMERGENCY Modes: .____1........1 .__I_2_.I I 3 ENERGY 1117/2018 Power Operations Startup Hot Shutdown NORTHWEST HOT CONDITIONS (RCS GT 200°F)

Initials Date

1. Number: 13.1.1A

Title:

Classifying The Emergency-Technical Bases

, Major Rev: 033 1 - - - - - - - - - - - - - - - - - - ' - - - - - - - - - - - - - - - - , Minor Rev: N/A Page: 1 of 190 PCN#:

PLANT PROCEDURES MANUAL N/A Effective Date:

IIIIIII IIIII IIII IIIIII IIII IIIIII IIIIII Ill llll 13.1.1A 01/17/18

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 2 of 190 DESCRIPTION OF CHANGES

!i\~*J~im?~f1,gij(r~qt)tr~*****

• New Stack Monitor has been installed, new values for classification thresholds have been applied to Table 3 Effluent Monitor Classification Thresholds 155, 157, 159, 161, Table 3 - Effluent Monitor Classification Thresholds 178 Revised to delete PRM-RE-18 and PRM-RE-1C and add PRM-RE-11,12,13 Revised threshold values to updated values from Cale NE-02-09-12 5 Added Engineering Calculation NE 09-12 to References Section 141 Revised "RWCU-Fl-620" to correct EPN "LD-Fl-620" .

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency- Technical Bases Page: 3 of 190 TABLE OF CONTENTS 1.0 PURPOSE ............................................................................................................................... 4

2.0 REFERENCES

........................................................................................................................ 5 3.0 DISCUSSION ........................................................................................................................... 6 3.1 Background ............................................................................................................................... 6 3.2 Fission Product Barriers ........................................................................................................... 6 3.3 Emergency Classification Based on Fission Product Barrier Degradation ................................ 7 3.4 EAL Organization ...................................................................................................................... 8 3.5 Technical Bases Information .................................................................................................. 10 3.6 Mode Applicability .................................................................................................................. 11 3.7 Basis ...................................................................................................................................... 11 3.8 CGS Basis Reference(s) ........................................................................................................ 11 3.9 Operating Mode Applicability ................................................................................................. 11 3.10 Storage Operations ................................................................................................................ 12 4.0 PROCEDURE ......................................................................... :.............................................. 12 4.1 General Considerations ......................................................................................................... 12 4.2 Classification Methodology ..................................................................................................... 14 5.0 DEFINITIONS ........................................................................................................................ 18 6.0 ATTACHMENTS ....................................................................................................:............... 30 6.1 EAL Technical Bases ............................................................................................................. 31 6.2 Tables .................................................................................................................................. 180 6.3 Safe Operation & Shutdown Areas Table 9 Bases ............................................................... 186 6.4 Emergency Classification Chart Distribution ......................................................................... 190

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency-Technical Bases Page: 4 of 190 1.0 PURPOSE 1.1 This document provides an explanation and rationale for each Emergency Action Level (EAL) included in the EAL Upgrade Project for Columbia Generating Station (CGS). It should be used to provide historical documentation for future reference and serve as a training aid.

Decision-makers responsible for implementation of PPM 13.1.1, Classifying the Emergency, may (though not required) use this document as a technical reference in support of EAL interpretation.

1.2 The expectation is that emergency classifications are to be made as soon as conditions are present and recognizable for the classification, but within 15 minutes or less in all cases of conditions present. Use of this document for assistance is not intended to delay the emergency classification.

1.3 Because the information in a basis document can affect emergency classification decision-making (e.g., the Emergency Director refers to it during an event), the NRC staff expects that changes to the basis document will be evaluated in accordance with the provisions of 10 CFR 50.54(q). Additionally, some criteria/values in the CGS EALs and fission product barrier thresholds are drawn from plant AOPs and EOPs. The impact of any changes to those procedures on EAL bases must' be evaluated for screening in accordance with the provisions of 10 CFR 50.54(q). This Emergency Plan Implementing Procedure is identified by reference in the Emergency Plan. Changes to the EAL Scheme (Attachments 6.1) require an LDCN since it is part of the Emergency Plan .

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 5 of 190

2.0 REFERENCES

2.1 Developmental 2.1.1 NEI 99-01 Revision 6, Methodology for the Development of Emergency Action Levels for Non-Passive Reactors, ADAMS Accession Number ML12326A805 2.1.2 Technical Specifications Table 1.1-1 Modes 2.1.3 NSIR/DPR-ISG-01 Interim Staff Guidance, Emergency Planning for Nuclear Power Plants 2.1.4 10CFR 50.72 Immediate Notification Requirements for Operating Nuclear Power Reactors 2.1.5 RIS 2007-02 Clarification of NRC Guidance for Emergency Notifications During Quickly Changing Events, February 2, 2007 2.1.6 NUREG-1022 Event Reporting Guidelines: 10CFR50.72 and 50.73 2.1.7 Hl-2002444, Holtec International Final Safety Analysis Report for the HI-STORM 100 Cask System, USNRC Docket No. 72-1014, Chapter 7, Confinement 2.1.8 PPM 1.20.3, Outage Risk Management 2.1.9 10CFR 50. 73 License Event Report System 2.1.10 M570, General Arrangement Plan - El. 572 ft - 0 in. and El. 606 ft - 10 1/2 in. -

Reactor Building 2.1.11 Certificate of Compliance No. 1014 Appendix A Technical Specifications for the HI-STORM 100 Cask System Section 1.1 Definitions 2.1.12 SWP-PR0-03, Procedure Writer's Manual 2.1.13 CGS Physical Security Plan 2.1.14 CGS Graphics Plant Drawing 902118-P 2.1.15 Energy Northwest Columbia Generating Station Offsite Dose Calculation Manual, Amendment 52 2.1.16 Engineering Calculation No. NE-02-09-12 2.2 Implementing 2.2.1 PPM 13.1.1, Classifying the Emergency 2.2.2 EP-01 Emergency Plan

/

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 6 of 190 2.2.3 Columbia Generating Station NEI 99-01 Revision 6 EAL Comparison Matrix 3.0 DISCUSSION 3.1 Background 3.1.1 EALs are the plant-specific indications, conditions or instrument readings that are utilized to classify emergency conditions defined in the CGS Emergency Plan.

3.1.2 In 1992, the NRC endorsed NUMARC/NESP-007 "Methodology for Development of Emergency Action Levels" as an alternative to NUREG-0654 EAL guidance.

3.1.3 NEI 99-01 (NUMARC/NESP-007) Revisions 4 and 5 were subsequently issued for industry implementation. Enhancements over earlier revisions included:

• Consolidating the system malfunction initiating conditions and example emergency action levels which address conditions that may be postulated to occur during plant shutdown conditions.

• Initiating conditions and example emergency action levels that fully address conditions that may be postulated to occur at permanently Defueled Stations and Independent Spent Fuel Storage Installations (ISFSls) .

• Simplifying the fission product barrier EAL threshold for a Site Area Emergency.

3.1.4 Subsequently, Revision 6 of NEI 99-01 has been issued which incorporates resolutions to numerous implementation issues including the NRC EAL Frequently Asked Questions (FAQs). Using NEI 99-01 Revision 6, "Methodology for the Development of Emergency Action Levels for Non-Passive Reactors," November 2012 (ADAMS Accession Number ML12326A805) (ref. 2.1.1), CGS conducted an EAL implementation upgrade project that produced the EALs discussed herein.

3.2 Fission Product Barriers Fission product barrier thresholds represent threats to the defense in depth design concept that precludes the release of radioactive fission products to the environment. This concept relies on multiple physical barriers, any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment.

Many of the EALs derived from the NEI methodology are fission product barrier threshold based. That is, the conditions that define the EALs are based upon thresholds that represent the loss or potential loss of one or more of the three fission product barriers. "Loss" and "Potential Loss" signify the relative damage and threat of damage to the barrier. A "Loss" threshold means the barrier no longer assures containment of radioactive materials. A "Potential Loss" threshold implies an increased probability of barrier loss and decreased certainty of maintaining the barrier .

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 7 of 190 3.2.1 The primary fission product barriers are:

a. Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.

b. Reactor Coolant System (RCS): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping out to and including the isolation valves.

c. Containment (PC): The drywell, the wetwell, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves comprise the PC barrier. Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to either a Site Area Emergency or a General Emergency using the Fission Product Barrier table.

3.3 Emergency Classification Based on Fission Product Barrier Degradation The following criteria are the bases for event classification related to fission product barrier loss or potential loss:

3.3.1 Alert

• 3.3.2 Any loss or any potential loss of either Fuel Clad or RCS barrier Site Area Emergency:

Loss or potential loss of any two barriers 3.3.3 General Emergency:

Loss of any two barriers and loss or potential loss of the third barrier

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 8 of 190 3.4 EAL Organization 3.4.1 The CGS EAL scheme includes the following features:

a. Division of the EAL set into three broad groups:

1) EALs applicable under all plant operating modes - This group would be reviewed by the EAL-user any time emergency classification is considered.

2) EALs applicable only under hot operating modes - This group would only be reviewed by the EAL-user when the plant is in Hot Shutdown, Startup, or Power Operations mode.

3) EALs applicable only under cold operating modes - This group would only be reviewed by the EAL-user when the plant is in Cold Shutdown, Refuel or Defueled mode.

3.4.2 The purpose of the groups is to avoid review of hot condition EALs when the plant is in a cold condition and avoid review of cold condition EALs when the plant is in a hot condition. This approach significantly minimizes the total number of EALs that must be reviewed by the EAL-user for a given plant condition, reduces EAL-user reading burden and, thereby, speeds identification of the EAL that applies to the emergency .

• 3.4.3 3.4.4 Within each group, assignment of EALs to categories and subcategories:

Category and subcategory titles are selected to represent conditions that are operationally significant to the EAL-user. The CGS EAL categories are aligned to and represent the NEI 99-01"Recognition Categories". Subcategories are used in the CGS scheme as necessary to further divide the EALs of a category into logical sets of possible emergency classification thresholds. The CGS EAL categories and subcategories are listed in Table 3.4-1.

3.4.5 The primary tool for determining the emergency classification level is the EAL Classification Matrix. The user of the EAL Classification Matrix may (but is not required to) consult the EAL bases in order to obtain additional information concerning the EALs under classification consideration. The user should consult Section 4.0 and Attachment 6.1 of this document for such information .

Number: 13.1.1A Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 9 of 190 Table 3.4-1 EAL Groups, Categories and Subcategories EAL Group/Category EAL Subcategory I

Any Operating Mode:

R - Abnormal Rad Release / Rad 1 - Radiological Effluent Effluent 2- Irradiated Fuel Event 3 - Area Radiation Levels H - Hazards and Other Conditions 1 -Security Affecting Plant Safety 2 - Seismic Event 3 - Natural or Technological Hazard 4-Fire 5 - Hazardous Gas 6 - Control Room Evacuation 7 - Emergency Director Judgment E - Independent Spent Fuel Storage 1 - Confinement Boundary Installation (ISFSI)

• Hot Conditions:

M - System Malfunction 1 - Loss of Emergency AC Power 2 - Loss of Vital DC Power 3- Loss of Control Room Indications 4 - RCS Activity 5 - RCS Leakage 6 - RPS Failure 7 - Loss of Communications 8 - Hazardous Event Affecting Safety Systems F - Fission Product Barrier None Degradation Cold Conditions:

C - Cold Shutdown / Refuel System 1-RPV Level Malfunction 2 - Loss of Emergency AC Power 3- RCS Temperature 4 - Loss of Vital DC Power 5 - Loss of Communications 6 - Hazardous Event Affecting Safety Systems

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency-Technical Bases Page: 10 of 190 3.5 Technical Bases Information 3.5.1 EAL technical bases are provided in Attachment 6.1. A summary explanation of each category and subcategory is given at the beginning of the technical bases discussions of the EALs included in the category. For each EAL, the following information is provided:

a. Category Letter & Title

b. Subcategory Number & Title

c. Initiating Condition (IC) 3.5.2 Site-specific description of the generic IC given in NEI 99-01 Rev. 6.

a. EAL Identifier (enclosed in rectangle)

1) Each EAL is assigned a unique identifier to support accurate communication of the emergency classification to onsite and offsite personnel. Four characters define each EAL identifier:

a) First character (letter): Corresponds to the EAL category as described above (R, C, H, M, F or E) b) Second character (letter): The emergency classification (G, S, A or U)

G = General Emergency S = Site Area Emergency A= Alert U = Unusual Event c) Third character (number): Subcategory number within the given category. Subcategories are sequentially numbered beginning with the number one (1). If a category does not have a subcategory, this character is assigned the number one (1).

d) Fourth character (number): The numerical sequence of the EAL within the EAL subcategory. If the subcategory has only one EAL, it is given the number one (1).

2) Classification (enclosed in rectangle)

Unusual Event (U), Alert (A), Site Area Emergency (S) or General Emergency (G)

3) EAL (enclosed in rectangle)

Exact wording of the EAL as it appears in the EAL Classification Matrix

Number: 13.1.1A j Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 11 of 190

4) Notes Any notes applicable to the EAL

5) Tables 3.6 Mode Applicability One or more of the following plant operating conditions comprise the mode to which each EAL is applicable: 1 - Power Operations, 2 - Startup, 3 - Hot Shutdown, 4 - Cold Shutdown, 5 - Refuel, D - Defueled, or All. Additionally, unique to the ISFSI, Storage Operations.

(See Section 2.10 for operating mode definitions).

3.7 Basis A basis section that provides CGS-relevant information concerning the EAL as well as a description of the rationale for the EAL as provided in NEI 99-01 Rev. 6.

3.8 CGS Basis Reference(s)

Site-specific source documentation from which the EAL is derived .

• 3.9 Operating Mode Applicability 3.9.1 Power Operations Reactor mode switch is in RUN.

3.9.2 Startup The mode switch is in STARTUP/HOT STANDBY or REFUEL with all reactor vessel head closure bolts fully tensioned.

3.9.3 Hot Shutdown The mode switch is in SHUTDOWN, with all reactor vessel head closure bolts fully tensioned, and reactor coolant temperature is GT 200°F.

3.9.4 Cold Shutdown The mode switch is in SHUTDOWN, all reactor vessel head closure bolts are fully tensioned, and reactor coolant temperature is LE 200°F.

3.9.5 Refuel The mode switch is in REFUEL or SHUTDOWN and one or more reactor vessel head closure bolts less than fully tensioned .

Number: 13. 1. 1A I Use Category: REFERENCE Major Rev: 033 Mihor Rev: N/A

Title:

Classifying The Emergency- Technical Bases Page: 12of 190 3.9.6 Defueled All reactor fuel removed from RPV. (Full core off load during refueling or extended outage).

• 3.9.7 The plant operating mode that exists at the time that the event occurs (prior to any protective system or operator action being initiated in response to the condition) should be compared to the mode applicability of the EALs. If a lower or higher plant operating mode is reached before the emergency classification is made, the declaration shall be based on the mode that existed at the time the event occurred.

3.9.8 For events that occur in Cold Shutdown or Refuel, escalation is via EALs that have Cold Shutdown or Refuel for mode applicability, even if Hot Shutdown (or a higher mode) is entered during any subsequent heat-up. In particular, the fission product barrier EALs are applicable only to events that initiate in Hot Shutdown or higher.

3.9.9 The ISFSI related EAL EU1. 1 is applicable in the Storage Operations mode as defined in the Certificate of Compliance Appendix A Section 1. 1 Definitions (ref 2.1.12).

3.10 Storage Operations

• Storage operations include all licensed activities that are performed at the ISFSI while a Spent Fuel Storage Cask (SFSC) containing spent fuel is situated within the ISFSI perimeter.

Storage Operations does not include MPC transfer between the Transfer Cask and the Overpack which begins when the MPC is lifted off the HI-TRAC bottom lid and ends when the MPC is supported from beneath by the Overpack (or the reverse).

4.0 PROCEDURE 4.1 General Considerations When making an emergency classification, the Emergency Director must consider all information having a bearing on the proper assessment of an Initiating Condition (IC). This includes the Emergency Action Level (EAL) plus the associated Operating Mode Applicability, Notes, and the informing basis information. In the Recognition Category F matrices, EALs are based on loss or potential loss of Fission Product Barrier Thresholds.

4. 1. 1 Classification Timeliness NRC regulations require the licensee to establish and maintain the capability to assess, classify, and declare an emergency condition within 15 minutes after the availability of indications to plant operators that an emergency action level has been exceeded and to promptly declare the emergency condition as soon as possible following identification of the appropriate emergency classification level. The NRC staff has provided guidance on implementing this requirement in NSIR/DPR-ISG-01, "Interim Staff Guidance, Emergency Planning for Nuclear Power Plants" (ref. 2. 1.3) .

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 13 of 190 4.1.2 Valid Indications All emergency classification assessments shall be based upon valid indications, reports or conditions. A valid indication, report, or condition, is one that has been verified through appropriate means such that there is no doubt regarding indicator operability, condition existence, or report accuracy. For example, verification could be accomplished through an instrument channel check, response on related or redundant indicators, or direct observation by plant personnel.

An indication, report, or condition is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to indicator operability, the condition existence, or the report accuracy is removed. Implicit in this definition is the need for timely assessment. The validation of indications should be completed in a manner that supports timely emergency declaration.

4.1.3 Imminent Conditions For ICs and EALs that have a stipulated time duration (e.g., 15 minutes, 30 minutes, etc.), the Emergency Director should not wait until the applicable time has elapsed, but should declare the event as soon as it is determined that the condition has exceeded, or will likely exceed, the applicable time. If an ongoing radiological release is detected and the release start time is unknown, it should be assumed that the release duration specified in the IC/EA~ has been exceeded, absent data to the contrary.

4.1.4 Planned vs. Unplanned Events A planned work activity that results in an expected event or condition which meets or exceeds an EAL does not warrant an emergency declaration provided that: 1) the activity proceeds as planned, and 2) the plant remains within the limits imposed by the operating license. Such activities include planned work to test, manipulate, repair, maintain or modify a system or component. In these cases, the controls associated with the planning, preparation and execution of the work will ensure that complian~e is maintained with all aspects of the operating license provided that the activity proceeds and concludes as expected. Events or conditions of this type may be subject to the reporting requirements of 1O § CFR 50. 72 (ref. 2.1.4) .

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency -Technical Bases Page: 14of190 4.1.5 Classification Based on Analysis The assessment of some EALs is based on the results of analyses that are necessary to ascertain whether a specific EAL threshold has been exceeded (e.g.,

dose assessments, chemistry sampling, RCS leak rate calculation, etc.). For these EALs, the EAL wording or the associated basis discussion will identify the necessary analysis. In these cases, the 15-minute declaration period starts with the availability of the analysis results that show the threshold to be exceeded (i.e., this is the time that the EAL information is first available). The NRC expects licensees to establish the capability to initiate and complete EAL-related analyses within a reasonable period of time (e.g., maintain the necessary expertise on-shift).

\

4.1.6 Emergency Director Judgment While the EALs have been developed to address a full spectrum of possible events and conditions which may warrant emergency classification, a provision for classification based on operator/management experience and judgment is still necessary. The NEI 99-01 EAL scheme provides the Emergency Director with the ability to classify events and conditions based upon judgment using EALs that are consistent with the Emergency Classification Level (ECL) definitions (refer to Category H). The Emergency Director will need to determine if the effects or consequences of the event or condition reasonably meet or exceed a particular ECL definition. A similar provision is incorporated in the Fission Product Barrier Tables; judgment may be used to determine the status of a fission product barrier.

4.2 Classification Methodology To make an emergency classification, the user will compare an event or condition (i.e., the relevant plant indications and reports) to an EAL(s) and determine if the EAL has been met or exceeded. The evaluation of an EAL must be consistent with the related Operating Mode Applicability and Notes. If an EAL has been met or exceeded, the associated IC is likewise met, the emergency classification process "clock" starts, and the ECL must be declared in accordance with plant procedures no later than fifteen minutes after the process "clock" started.

When assessing an EAL that specifies a time duration for the off-normal condition, the "clock" for the EAL time duration runs concurrently with the emergency classification process "clock."

For a full discussion of this timing requirement, refer to NSIR/DPR-ISG-01 (ref. 2.1.3).

4.2.1 Classification of Multiple Events and Conditions

a. When multiple emergency events or conditions are present, the user will identify all met or exceeded EALs. The highest applicable ECL identified during this review is declared. For example:

• If an Alert EAL and a Site Area Emergency EAL are met, a Site Area Emergency should be declared .

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 15 of 190

b. There is no "additive" effect from multiple EALs meeting the same ECL. For example:

• If two Alert EALs are met, an Alert should be declared.

c. Related guidance concerning classification of rapidly escalating events or conditions is provided in Regulatory Issue Summary (RIS) 2007-02, Clarification of NRG Guidance for Emergency Notifications During Quickly Changing Events (ref. 2.1.5).

4.2.2 Consideration of Mode Changes During Classification

a. The mode in effect at the time that an event or condition occurred, and prior to any plant or operator response, is the mode that determines whether or not an IC is applicable. If an event or condition occurs, and results in a mode change before the emergency is declared, the emergency classification level is still based on the mode that existed at the time that the event or condition was initiated (and not when it was declared). Once a different mode is reached, any new event or condition, not related to the original event or condition, requiring emergency classification should be evaluated against the ICs and EALs applicable to the operating mode at the time of the new event or condition.

• b. For events that occur in Cold Shutdown or Refuel, escalation is via EALs that are applicable in the Cold Shutdown or Refuel modes, even if Hot Shutdown (or a higher mode) is entered during the subsequent plant response. In particular, the fission product barrier EALs are applicable only to events that initiate in the Hot Shutdown mode or higher.

4.2.3 Classification of Imminent Conditions Although EALs provide specific thresholds, the Emergency Director must remain alert to events or conditions that could lead to meeting or exceeding an EAL within a relatively short period of time (i.e., a change in the ECL is IMMINENT). If, in the judgment of the Emergency Director, meeting an EAL is IMMINENT, the emergency classification should be made as if the EAL has been met. While applicable to all emergency classification levels, this approach is particularly important at the higher emergency classification levels since it provides additional time for implementation of protective measures.

4.2.4 Emergency Classification Level Upgrading and Downgrading

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

b. As noted above, guidance concerning classification of rapidly escalating events or conditions is provided in RIS 2007-02 (ref. 2.1.5).

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 16 of 190 4.2.5 Classification of Short-Lived Events Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance. By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed. If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration.

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

4.2.6 Classification of Transient Conditions Many of the ICs and/or EALs employ time-based criteria. These criteria will require that the IC/EAL conditions be present for a defined period of time before an emergency declaration is warranted. In cases where no time-based criterion is specified, it is recognized that some transient conditions may cause an EAL to be met for a brief period of time (e.g., a few seconds to a few minutes). The following guidance should be applied to the classification of these conditions.

a. EAL momentarily met during expected plant response - In instances where an EAL is briefly met during an expected (normal) plant response, an emergency declaration is not warranted provided that associated systems and components are operating as expected, and operator actions are performed in accordance with procedures.

b. EAL momentarily met but the condition is corrected prior to an emergency declaration - If an operator takes prompt manual action to address a condition, and the action is successful in correcting the condition prior to the emergency declaration, then the applicable EAL is not considered met and the associated emergency declaration is not required. For illustrative purposes, consider the following example:

An ATWS occurs and the high pressure ECCS systems fail to automatically start. RPV level rapidly decreases and the plant enters an inadequate core cooling condition (a potential loss of both the fuel clad and RCS barriers). If an operator manually starts a high pressure ECCS system in accordance with an EOP step and clears the inadequate core cooling condition prior to an emergency declaration, then the classification should be based on the ATWS only .

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency -Technical Bases Page: 17 of 190

c. It is important to stress that the 15-minute emergency classification assessment period (process clock) is not a "grace period" during which a classification may be delayed to allow the performance of a corrective action that would obviate the need to classify the event. Emergency classification assessments must be deliberate and timely, with no undue delays. The provision discussed above addresses only those rapidly evolving situations when an operator is able to take a successful corrective action prior to the Emergency Director completing the review and steps necessary to make the emergency declaration. This provision is included to ensure that any public protective actions resulting from the emergency classification are truly warranted by. the plant conditions.

4.2. 7 Transitory Event Classification

a. In some cases, an EAL may be met but the emergency classification was not made at the time of the event or condition. This situation can occur when personnel discover that an event or condition existed which met an EAL, but no emergency was declared, and the event or condition no longer exists at the time of discovery. This may be due to the event or condition not being recognized at the time or an error that was made in the emergency classification process.

b. In these cases, no emergency declaration is warranted; however, the guidance contained in NUREG-1022 (ref. 2.1.6) is applicable. Specifically, the event should be reported to the NRC in accordance with 10 CFR § 50. 72 (ref. 2.1.5) within one hour of the discovery of the undeclared event or condition. The licensee should also notify appropriate State and local agencies in accordance with the agreed upon arrangements.

In some cases, the licensee discovers that a condition existed that met the Emergency Action Level (EAL) criteria, but no emergency was declared and the basis for the emergency classification no longer exists at the time of this discovery. This may be due to a rapidly concluded event or an oversight in the emergency classification made during the event, or it may be determined during a post event review. In these cases, in accordance with NUREG 1022, no emergency declaration is warranted.

If the licensee does not declare an emergency under these circumstances, an Emergency Notification System (ENS) notificatio_n within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of the discovery of the undeclared (or misclassified) event should be performed, in accordance with PPM 13.4.1.

If the licensee does declare an emergency, then all notifications required by PPM 13.4.1 are to be made.

4.2.8 Retraction of an Emergency Declaration Guidance on the retraction of an emergency declaration reported to the NRC is discussed in NUREG-1022 (ref. 2.1.6).

Number: 13.1.1A

, \ Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 18 of 190 5.0 DEFINITIONS 5.1 Definitions (ref. 2.1.1 except as noted)

Selected terms used in Initiating Condition and Emergency Action Level statements are set in all capital letters (e.g., ALL CAPS). These words a*re defined terms that have specific meanings as used in this document. The definitions of these terms are provided below.

5.1.1 ALERT Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA PAG exposure levels.

5.1.2 CAN/CANNOT BE MAINTAINED ABOVE/BELOW The value of an identified parameter is/is not able to be held within the specified limit. The determination requires an evaluation of system performance and availability in relation to parameter values and trends. An instruction prescribing action when a parameter cannot be maintained above or below a specified limit neither requires nor prohibits anticipatory action-depending upon plant conditions, the action may be taken as soon as it is determined that the limit will ultimately be exceeded, or delayed until the limit is actually reached. Once the parameter does exceed the limit, however, the action must be performed; it may not be delayed while attempts are made to restore the parameter to within the desired control band.

5.1.3 CAN/CANNOT BE RESTORED ABOVE/BELOW The value of an identified parameter is/is not able to be brought within the specified limit. The determination requires an evaluation of system performance and availability in relation to parameter values and trends. An instruction prescribing action when a value cannot be restored and maintained above or below a specified limit does not require immediate action simply because the current values is outside the range, but does not permit extended operation beyond the limit; the action must be taken as soon as it is apparent that the specified range cannot be attained.

5.1.4 CONFINEMENT BOUNDARY The barrier(s) between spent fuel and the environment once the spent fuel is processed for dry storage. As related to the CGS ISFSI, Confinement Boundary is defined as the Multi-Purpose Canister (MPC) (ref. 2.1.7) .

Number: 13.1.1A J Use Category: REFERENCE Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 19 of 190 5.1.5 CONTAINMENT CLOSURE The procedurally defined conditions or actions taken to secure Containment (Primary or Secondary) and its associated structures, systems, and components as a functional barrier to fission product release under shutdown conditions. A functional barrier is one which mitigates offsite release during an event.

Containment Closure requires a functional barrier (not necessarily Technical Specification Operable; the appropriate structures, systems, and components are functional) to exist at the time of an event. The site cannot rely on contingency methods to establish a functional barrier after the event has started. In Mode 4 either a functional Primary Containment or a functional Secondary Containment is sufficient to mitigate offsite release. In Mode 5, a functional Secondary Containment is sufficient to mitigate offsite release. Therefore, Containment Closure is met in Mode 4 with either a functional Primary Containment or a functional Secondary Containment. Containment Closure is met in Mode 5 with a functional Secondary Containment.

5.1.6 EPA PAGS Environmental Protection Agency Protective Action Guidelines. The EPA PAGs are expressed in terms of dose commitment: 1 Rem TEDE or 5 Rem COE Thyroid.

Actual or projected offsite exposures in excess of the EPA PAGs requires CGS to recommend protective actions for the general public to Offsite Response Organizations (ORO).

5.1.7 EMERGENCY ACTION LEVEL A pre-determined, site-specific, observable threshold for an Initiating Condition that, when met or exceeded, places the plant in a given emergency classification level.

5.1.8 EMERGENCY CLASSIFICATION LEVEL One of a set of names or titles established by the US Nuclear Regulatory Commission (NRC) for grouping off-normal events or conditions according to (1) potential or actual effects or consequences, and (2) resulting onsite and offsite response actions. The emergency classification levels, in ascending order of severity, are:

• Unusual Event (UE)

• Alert

• Site Area Emergency (SAE)

• General Emergency

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency- Technical Bases Page: 20 of 190 5.1.9 EXPLOSION A rapid, violent and catastrophic failure of a piece of equipment due to combustion, chemical reaction or overpressurization. A release of steam (from high energy lines or components) or an electrical component failure (caused by short circuits, grounding, arcing, etc.) should not automatically be considered an explosion. Such events require a post-event inspection to determine if the attributes of an explosion are present.

5.1.10 FIRE Combustion characterized by heat and light. Sources of smoke such as slipping drive belts or overheated electrical equipment do not constitute fires. Observation of flame is preferred but is NOT required if large quantities of smoke and heat are observed.

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

5.1.12 FLOODING 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.

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

5.1.14 HOSTAGE A person(s) held as leverage against the station to ensure that demands will be met by the station .

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 21 of 190 5.1.15 HOSTILE ACTION An act toward CGS or its personnel that includes the use of violent force to destroy equipment, take hostages, and/or intimidate Energy Northwest to achieve an end.

This includes attack by air, land, or water using guns, explosives, projectiles, vehicles, or other devices used to deliver destructive force. Other acts that satisfy the overall intent may be included. Hostile action should not be construed to include acts of civil disobedience or felonious acts that are not part of a concerted attack on CGS. Non-terrorism-based EALs should be used to address such activities (i.e., this may include violent acts between individuals in the Owner Controlled Area).

5.1.16 HOSTILE FORCE One or more individuals who are engaged in a determined assault, overtly or by stealth and deception, equipped with suitable weapons capable of killing, maiming, or causing destruction.

5.1.17 IMMINENT The trajectory of events or conditions is such that an EAL will be met within a relatively short period of time regardless of mitigation or corrective actions .

• 5.1.18 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).

5.1.19 INDEPENDENT SPENT FUEL STORAGE INSTALLATION (ISFSI)

A complex that is designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage.

5.1.20 INITIATING CONDITION An event or condition that aligns with the definition of one of the four emergency classification levels by virtue of the potential or actual effects or consequences.

5.1.21 MAINTAIN Take appropriate action to hold the value of an identified parameter within specified limits.

5.1.22 NORMAL LEVELS As applied to radiological IC/EALs, the highest reading in the past twenty-four hours excluding the current peak value.

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 22 of 190 5.1.23 OWNER CONTROLLED AREA The area that Energy Northwest maintains industrial and process control of (ref. 2.2.2).

5.1.24 PROJECTILE An object directed toward CGS that could cause concern for its continued operability, reliability, or personnel safety.

5.1.25 PROTECTED AREA An area located within the OWNER CONTROLLED AREA which contains the Columbia Generating Station power block and is surrounded by chain link fence (ref. 2.2.2).

5.1.26 RCS INTACT The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams) .

• 5.1.27 5.1.28 REFUELING PATHWAY Reactor cavity and spent fuel pool comprise the Refuel Pathway (ref. 2.1.11 ).

SAFETY SYSTEM A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. These are typically systems classified as safety-related (as defined in 10CFR50.2):

I Those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

a. The integrity of the reactor coolant pressure boundary;

b. The capability to shut down the reactor and maintain it in a safe shutdown condition;

c. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures .

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 23 of 190 5.1.29 SECURITY CONDITION Any security event as listed in the approved security contingency plan that constitutes a threat/compromise to site security, threat/risk to site personnel, or a potential degradation to the level of safety of the plant. A security condition does not involve a HOSTILE ACTION.

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

5.1.31 SITE BOUNDARY 1950-meter radius around the plant as depicted in Figure 3-1 of the CGS ODCM (ref. 2.1.16).r The key-hole area between the river and this radius is not within the Site Boundary .

5.1.32 UNISOLABLE An open or breached system line that cannot be isolated, remotely or locally.

5.1.33 UNPLANNED.

A parameter change or an event that is not: 1) the result of an intended evolution, or

2) an expected plant response to a transient. The cause of the parameter change or event may be known or unknown.

5.1.34 UNUSUAL EVENT Events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

/'

5.1.35 VALID An indication, report, or condition, is considered to be valid when it is verified by (1) an instrument channel check, or (2) indications on related or redundant indicators, or (3) by direct observation by plant personnel, such that doubt related to the indicator's operability, the condition's existence, or the report's accuracy is removed.

Implicit in this definition is the need for timely assessment.

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 24 of 190 I

5.1.36 VISIBLE DAMAGE Damage to a SAFETY SYSTEM train that is readily observable without measurements, testing, or analysis. The visual impact of the damage is sufficient to cause concern regarding the operability or reliability of the affected SAFETY SYSTEM train.

5.2 Abbreviations/Acronyms OF Degrees Fahrenheit 0

Degrees AC Alternating Current APRM Average Power Range Meter ARI Automatic Rod Insertion ATWS Anticipated Transient Without Scram BWR Boiling Water Reactor BWROG Boiling Water Reactor Owners Group COE Committed Dose Equivalent CFR

• Code of Federal Regulations cps counts per second OBA Design Basis Accident DC Direct Current EAL Emergency Action Level ECCS Emergency Core Cooling System ECL Emergency Classification Level EOF Emergency Operations Facility EOP Emergency Operating Procedure EPA Environmental Protection Agency EPG Emergency Procedure Guideline EPIP Emergency Plan Implementing Procedure ESF Engineered Safety Feature FAA Federal Aviation Administration FBI Federal Bureau of Investigation FEMA Federal Emergency Management Agency FSAR Final Safety Analysis Report GOS *Graphic Display System

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 25 of 190 GE Greater than or Equal to gm Gram GT Greater Than HCTL Heat Capacity Temperature Limit HPCS High Pressure Core Spray HOO NRC Headquarters Operations Officer IC Initiating Condition ISFSI Independent Spent Fuel Storage ln~tallation Kett Effective Neutron Multiplication Factor LCO Limiting Condition of Operation LE Less than or Equal to LOCA Loss of Coolant Accident LPCS Low Pressure Core Spray LT Less Than MPC Maximum Permissible Concentration/Multi-Purpose Canister

µCi Micro Curie MSCRWL Minimum Steam Cooling RPV Water Level MSIV Main Steam Isolation Valve MSL Main Steam Line mR milliRoentgen MW Megawatt NEI Nuclear Energy Institute NESP National Environmental Studies Project NORAD North American Aerospace Defense Command NRC Nuclear Regulatory Commission NSSS Nuclear Steam Supply System OBE Operating Basis Earthquake OCA Owner Controlled Area ODCM Off-site Dose Calculation Manual ORO Offsite Response Organization PMU Panel Meter Unit PRA Probabilistic Risk Assessment

• PRM Process Radiation Monitor

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 26 of 190 PSIG Pounds per Square Inch Gauge PSP Pressure Suppression Pressure R Roentgen RB Reactor Building RCC Reactor Building Closed Cooling RCIC Reactor Core Isolation Cooling RCS Reactor Coolant System Rem Roentgen Equivalent Man RHR Residual Heat Removal RPS Reactor Protection System RPV Reactor Pressure Vessel RWCU Reactor Water Cleanup SGT Stand-By Gas Treatment SBO Station Blackout SDSP Shutdown Safety Plan SLC Standby Liquid Control SPDS Safety Parameter Display System SRO Senior Reactor Operator SSC Structure, System or Component SW Service Water TEA Turbine Exhaust Air TEDE Total Effective Dose Equivalent TAF Top of Active Fuel TSC Technical Support Center TSW Plant Service Water WEA Waste Exhaust Air

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 27 of 190 5.3 CGS-TO-NEI 99-01 Rev. 6 EAL CROSS-'REFERENCE This cross-reference is provided to facilitate association and location of a CGS EAL within the NEI 99-01 IC/EAL identification scheme. Further information regarding the development of the CGS EALs based on the NEI guidance can be found in the EAL Comparison Matrix.

NEI 99-01 Rev. 6 CGS EAL Example IC EAL CG1.1 CG1 1 CS1.1 CS1 1, 2 CA1.1 CA1 1, 2 CU1.1 CU1 1, 2 CG1.2 CG1 2 CS1.2 CS1 3

• CA2.1 CU2.1 CA3.1 CU3.1 CA2 CU2 CA3 CU3 1

1 1, 2 1

CU3.2 CU3 2 CU4.1 CU4 1 CUS.1 cus 1, 2, 3 CA6.1 CA6 1 EU1.1 E-HU1 1 FG1.1 FG1 1 FS1.1 FS1 1 FA1.1 FA1 1 HS1.1 HS1 1 HA1.1 HA1 1, 2

• HU1.1 HU1 1, 2 3

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 28 of 190 NEI 99-01 Rev. 6 CGS EAL Example IC EAL HU2.1 HU2 1 HU3.1 HU3 1, 5 HU3.2 HU3 2 HU3.3 HU3 3,4 HU4.1 HU4 1 HU4.2 HU4 2 HU4.3 HU4 3,4 HA5.1 HAS 1 HS6.1 HS6 1 HA6.1 HA6 1 HG7.1 HG? 1 HS7.1 HS7 1 HA7.1 HA? 1 HU7.1 HU? 1 MG1.1 SG1 1 MS1.1 SS1 1 MA1.1 SA1 1 MU1.1 SU1 1 MG1.2 SGS 1 MS2.1 SS8 1 MA3.1 SA2 1 MU3.1 SU2 1 MU4.1 SU3 1

• MU4.2 SU3 2

IUse Category: REFERENCE Number: 13.1.1A Major Rev:

• 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 29 of 190 NEI 99-01 Rev. 6 CGS EAL Example IC EAL MUS.1 SU4 1, 2, 3 MS6.1 SSS 1 MA6.1 SAS 1 MU6.1 SUS 1, 2 MU7.1 SU6 1, 2, 3 MA8.1 SA9 1 RG1.1 AG1 1, 2 RS1.1 AS1 1, 2 RA1.1 AA1 1, 2 RU1.1 AU1 1, 2, 3 RG1.2 AG1 3 RS1.2 AS1 3 RA1.2 AA1 3 RA1.3 AA1 4 RG2.1 AG2 1 RS2.1 AS2 1 RA2.1 AA2 1 RU2.1 AU2 1 RA2.2 AA2 2 RA2.3 AA2 3 RA3.1 AA3 1, 2

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 30 of 190 6.0 ATTACHMENTS 6.1 EAL Technical Bases 6.2 Tables and Notes 6.3 Table 9 Basis 6.4 Emergency Action Level Chart Distribution

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 31 of 190 EAL TECHNICAL BASES Category C - Cold Shutdown / Refuel System Malfunction EAL Group: Cold Conditions (RCS temperature s 200°F);

EALs in this category are applicable only in one or more cold operating modes.

Category C EALs are directly associated with cold shutdown or Refuel system safety functions. Given the variability of plant configurations (e.g., systems out-of-service for maintenance, containment open, reduced AC power redundancy, time since shutdown) during these periods, the consequences of any given initiating event can vary greatly. For example, a loss of decay heat removal capability that occurs at the end of an extended outage has less significance than a similar loss occurring during the first week after shutdown. Compounding these events is the likelihood that instrumentation necessary for assessment may also be inoperable. The cold shutdown and Refuel system malfunction EALs are based on performance capability to the extent possible with consideration given to RCS integrity, CONTAINMENT CLOSURE, and fuel clad integrity for the applicable operating modes (4 - Cold Shutdown, 5 - Refuel, D- Defueled).

The events of this category pertain to the following subcategories:

1. RPV Level Reactor Pressure V.essel water level is directly related to the status of adequate core cooling and, therefore, fuel clad integrity .

• 2. Loss of Emergency AC Power Loss of emergency plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite power sources for 4160 V emergency buses.

3. RCS Temperature Uncontrolled or inadvertent temperature or pressure increases are indicative of a potential loss of safety functions.

4. Loss of Vital DC Power Loss of vital plant electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the vital125 voe buses.

5. 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.

6. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification .

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 32 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 1-RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged EAL:

CG1 .1 General Emergency RPV level LT-161 in. for GE 30 min. (Note 1)

AND Any of the following indications of Containment Challenge:

• CONTAINMENT CLOSURE not established (Note 6)

• Explosive mixture inside PC (H 2 GE 6% and 0 2 GE 5%)

• UNPLANNED rise in PC pressure

• RB area radiation GT any Maximum Safe Operating level (PPM 5.3.1 Table 24)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded

• Mode Applicability:

Basis:

When RPV level drops to the top of active fuel (an indicated RPV level of -161 in.), core uncovery starts to occur (ref. 1, 2).

Four conditions are associated with a challenge to primary containment (PC) integrity:

• Containment Closure is defined as the Shutdown Safety Plan (SDSP) actions taken to secure primary or secondary containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions. This definition is less restrictive than Technical Specification criteria governing Primary and Secondary Containment operability. If the Technical Specification criteria are met, therefore, Containment Closure has been established. (ref. 3, 4, 5)

• Explosive (deflagration) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration

• concentration limit (ref. 6) .

Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency-Technical Bases Page: 33 of 190 The specified values for this threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref. 5) and readily recognizable because 6% hydrogen is well above the EOP flowchart entry condition (ref. 8). The minimum global deflagration hydrogen/oxygen concentrations (6% / 5%, respectively) require intentional primary containment venting, which is defined to be a loss of the primary containment barrier.

Atmosphere samples from a minimum of two locations inside the primary containment and one location in the suppression chamber are sequentially monitored for hydrogen and oxygen percentage levels by each of two redundant analyzer systems. The analyzers are single range (0 to 30% hydrogen and Oto 30% oxygen). Two redundant (divisional) recorders are provided in the Main Control Room CMS-02/H2R-1 (H13-P827) and CMS-02/H2R-2 (H13-P811).

Hydrogen and oxygen concentrations can also be displayed on the plant computers (ref. 9-12)

• Any UNPLANNED rise in PC pressure in the Cold Shutdown or Refueling mode indicates Containment Closure cannot be assured and the primary containment cannot be relied upon as a barrier to fission product release.

• RB (Reactor Building) area radiation monitors should provide indication of increased release that may be indicative of a challenge to Containment Closure. The Maximum Safe Operating levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in Table 24 of the EOP flowcharts (ref. 13). All Table 24 Maximum Safe Operating radiation levels can be determined in the main Control Room .

• If RPV level is restored and maintained above the top of active fuel before a Containment Challenge condition occurs and subsequently a Containment Challenge condition is reached, this EAL is not met.

This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable.

With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is

• challenged .

Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 34 of 190 The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operations at Commercial Nuclear Power Plants in the United States; and NU MARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

CGS Basis Reference(s):

1. Calculation NE-02-03-05 Attachment 3 Note 8

2. PPM 5.1.1 RPV Control

3. Technical Specifications 3.6.1.1

4. Technical Specifications 3.6.4.1

5. PPM 1.20.3 Outage Risk Management

6. BWROG EPG/SAG Revision 2, Sections PC/G

7. PPM 5.7.1 RPV and Primary Containment Flooding SAG, Table 19

8. PPM 5.2.1 Primary Containment Control

9. FSAR Section 7.5.1.5.4

10. PPM 5.0.10 Flowchart Training Manual

11. PPM 4.814.J1 814.J1 Annunciator Panel Alarms, 2-2

12. PPM 4.814.J2 814.J2 Annunciator Panel Alarms, 2-2

13. PPM 5.3.1 Secondary Containment Control

14. NEI 99-01 CG1

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency -Technical Bases Page: 35 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS1 .1 Site Area Emergency (1) CONTAINMENT CLOSURE not established AND RPV level LT-129 in.

OR (2) CONTAINMENT CLOSURE established AND RPV level LT-161 in.

Mode Applicability:

Basis:

EAL#1 The threshold RPV water level of-129 in. is the low-low-low ECCS actuation setpoint. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further RPV water level decrease and potential core uncovery. The inability to restore and maintain level after reaching this setpoint infers a failure of the RCS barrier and Potential Loss of the Fuel Clad barrier. (ref. 1)

EAL#2 When RPV level drops to the top of active fuel (an indicated RPV level of -161 in.), core uncovery starts to occur (ref. 2).

This IC addresses a significant and prolonged loss of RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable.

Outage/shutdown contingency plans typically provide for re-establishing or verifying CONTAINMENT CLOSURE following a loss of heat removal or RCS inventory control functions. The difference in the specified RPV levels of CS 1.1 (1) and CS1 .1 (2) reflect the fact that with CONTAl NM ENT CLOSURE established, there is a lower probability of a fission product release to the environment.

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The l;:mergency - Technical Bases Page: 36 of 190 This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operations at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG1 or RG1.

CGS Basis Reference(s):

1. Technicpl Specifications Table 3.3.5.1-1, "Emergency Core Cooling System Instrumentation"

2. PPM 5.1.1 RPV Control

3. NEI 99-01 CS1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 37 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Significant loss of RPV inventory EAL:

CA1.1 Alert (1) Loss of RPV inventory as indicated by RPV level LT -50 in.

OR (2) RPV level cannot be monitored for GE 15 min. (Note 1)

AND UNPLANNED increase in any Table 1 sump or pool levels due to a loss of RPV inventory Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 1 Sumps/Pool

• Any valid Hi-Hi level alarm on R-1 through R-5 sumps

• EDR GE 25 GPM

• FDR GE 10 GPM

• Wetwell level rise

• Observation of UNISOLABLE RCS leakage Mode Applicability:

Basis:

This IC addresses conditions that are precursors to a loss of the ability to adequately cool irradiated fuel (i.e., a precursor to a challenge to the fuel clad barrier). This condition represents a potential substantial reduction in the level of plant safety.

If RPV water level continues to lower, then escalation to Site Area Emergency would be via IC CS1.

EAL#1 The threshold RPV level of -50 in. is the low-low ECCS (HPCS) actuation setpoint (ref. 1, 2).

In Cold Shutdown mode, the RCS will normally be intact and standard RPV level monitoring means are available. In the Refuel mode, the RCS is not intact and RPV level may be monitored by different means, including the ability to monitor level visually .

• Attachment 6.1, EAL Technical .Bases

IUse Category: REFER.ENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 38 of 190 For EAL #1, a lowering of water level below -50 in. indicates that operator actions have not been successful in restoring and maintaining RPV water level. The heat-up rate of the coolant will increase as the available water inventory is reduced. A continuing decrease in water level will lead to core uncovery.

Although related, EAL #1 is concerned with the loss of RCS inventory and not the potential concurrent effects on systems needed for decay heat removal (e.g., loss of a Decay Heat Removal suction point).

An increase in RCS temperature caused by a loss of decay heat removal capability is evaluated under ICCA3.

EAL#2 In this EAL, all RPV level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table 1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Reactor Building equipment or floor drain sump level rise may be indicative of RPV inventory losses external to the primary containment from systems connected to the RPV (ref. 3, 4).

With RHR System operating in the Shutdown Cooling mode, an unexplained rise in wetwell level could be indicative of RHR valve misalignment or leakage (ref. 5). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the primary containment that cannot be isolated could be indicative of a loss of RPV inventory .

• For EAL #2, the inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank l~vels. Sump and/or tank level changes must be evaluated against other potential sources of water flow'to ensure they are indicative of leakage from the RPV.

The 15-minute duration for the loss of level indication was chosen because it is half of the EAL duration specified in IC CS1.

CGS Basis Reference(s):

1. Technical Specifications Table 3.3.5.1-1

2. PPM 5.1.1 RPV Control

3. SOP-EDR-OPS Equipment Drain System Operation

4. SOP-FDR-OPS Floor Drain System Operation

5. SOP-RHR-SDC RHR Shutdown Cooling

6. NEI 99-01 CA 1

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 39 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 1-RPV Level Initiating Condition: UNPLANNED loss of RPV inventory EAL:

CU1.1 Unusual Event (1) UNPLANNED loss of reactor coolant results in RPV level less than a required lower limit for GE 15 min. (Note 1)

OR (2) RPV level cannot be monitored AND UNPLANNED increase in any Table 1 sump or pool levels due to a loss of RPV inventory Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 1 Sumps/Pool

• Any valid Hi-Hi level alarm on R-1 through R-5 sumps EDR GE 25 GPM FDR GE 10 GPM Wetwell level rise

• Observation of UNISOLABLE RCS leakage Mode Applicability:

Basis:

These Cold Shutdown EALs represent the hot condition EAL MU5.1, in which RCS leakage is associated with Technical Specification limits. In Cold Shutdown, these limits are not applicable; hence, the use of RPV level as the parameter of concern in this EAL.

This IC addresses the inability to restore and maintain water level to a required minimum level (or the ,

lower limit of a level band), or a loss of the ability to monitor RPV level concurrent with indications of coolant leakage. Either of these conditions is considered to be a potential degradation of the level of safety of the plant.

Refuel evolutions that decrease RCS water inventory are carefully planned and controlled. An UNPLANNED event that results in water level decreasing below a procedurally required limit warrants the declaration of an Unusual Event due to the reduced water inventory that is available to keep the

• core covered .

Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 40 of 190 Continued loss of RCS inventory may result in escalation to the Alert emergency classification level via either IC CA1 or CA3.

EAL#1 In Mode 4 and Mode 5, prior to flood up, RPV level is monitored from -310 in. to +400 in. to ensure adequate coverage for expected and postulated conditions of RPV level. All instr~ments are referenced to a benchmark at 527 .5 in. above the inside bottom head of the reactor vessel. This benchmark corresponds to the bottom edge of the steam dryer skirt and is the O in. reference indication on the RPV level instruments (ref. 1, 2, 3).

In preparation for refueling operations, level instruments are modified to provide continuous level indication from within the RPV to the refuel floor (ref. 4, 5).

The RPV level is controlled in a designated band in the reactor vessel and it is the inability to maintain level above the low end of the designated control band due to a loss of inventory resulting from a leak in the RCS that is the concern. With the plant in Refuel mode, RPV water level is normally maintained at or above the reactor vessel flange (ref. 6).

EAL #1 recognizes that the minimum required RPV level can change several times during the course of a Refuel outage as different plant configurations and system lineups are implemented. This EAL is met if the minimum level, specified for the current plant conditions, cannot be maintained for 15 minutes or longer. The minimum level is typically specified in the applicable operating procedure but may be specified in another controlling document.

• The 15-minute threshold duration allows sufficient time for prompt operator actions to restore and maintain the expected water level. This criterion excludes transient conditions causing a brief lowering of water level.

EAL#2 In this EAL, all RPV level indication is unavailable and the RPV inventory loss must be detected by the leakage indications listed in Table 1. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage. Reactor Building equipment or floor drain sump level rise may be indicative of RPV inventory losses external to the primary containment from systems connected to the RPV (ref. 7, 8, 6).

With RHR System operating in the Shutdown Cooling mode, an unexplained rise in wetwell level could be indicative of RHR valve misalignment or leakage (ref. 10). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified. Visual observation of leakage from systems connected to the RCS in areas outside the primary containment that cannot be isolated could be indicative of a loss of RPV inventory.

EAL #2 addresses a condition where all means to determine RPV level have been lost. In this condition, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RPV .

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 41 of 190 CGS Basis Reference(s):

1. FSAR Section 7.5.1.1

2. FSAR Table 7.5-1

3. FSAR Figure 7.7-1

4. PPM 10.27.39 Refueling Reactor Vessel Level (Temporary)

5. SOP-CAVITY-FILL Reactor Cavity and Dryer Separator Pit Fill

6. Technical Specifications 3.9.6

7. FSAR Section 7.6.1.3

8. SOP-EDR-OPS Equipment Drain System Operation

9. SOP-FDR-OPS Floor Drain System Operation

10. SOP-RHR-SDC RHR Shutdown Cooling

11. NEI 99-01 CU1

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 42 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting fuel clad integrity with containment challenged EAL:

CG1 .2 General Emergency RPV level cannot be monitored for GE 30 min. (Note 1)

AND Core uncovery is indicated by any of the following:

• UNPLANNED wetwell level rise GT 2 inches (PPM 5.2.1 entry condition)

• Valid indication of RB room flooding as identified by high level alarms (PPM 5.3.1 Table 25)

• Observation of UNISOLABLE RCS leakage outside primary containment of sufficient magnitude to indicate core uncover AND Any of the following indication of containment challenge:

CONTAINMENT CLOSURE not established (Note 6)

Explosive mixture inside PC (H2 GE 6% and 02 GE 5%)

UNPLANNED rise in PC pressure RB area radiation GT any Maximum Safe Operating level (PPM 5.3.1 Table 24)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required Mode Applicability:

Basis:

In this EAL, all RPV level indication is unavailable and the RPV inventory loss must be detected by the leakage indications provided. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage.

Reactor Building equipment or floor drain sump level rise may be indicative of RPV inventory losses external to the primary containment from systems connected to the RPV (ref. 1, 2). With RHR System operating in the Shutdown Cooling mqde, an unexplained rise in wetwell level could be indicative of RHR valve misalignment or leakage (ref. 3). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage

• cannot be immediately identified .

Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency- Technical Bases Page: 43 of 190 An UNPLANNED wetwell level increase to GT 2 inches or a VALID RB room high level alarm indicates a significant loss of RCS that could lead to core uncovery if not isolated (ref. 4, 5).

Visual observation of significant leakage from systems connected to the RCS in areas outside the primary containment that cannot be isolated could be indicative of a loss of RPV inventory sufficient to lead to core uncovery.

Four conditions are associated with a challenge to primary containment (PC) integrity:

• CONTAINMENT CLOSURE is not established.

• Explosive (deflagration) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref.

6).

• The specified values for this threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref. 6) and readily recognizable because 6% hydrogen is well above the EOP flowchart entry condition (ref. 8). The minimum global deflagration hydrogen/oxygen concentrations (6%/5%, respectively) require intentional primary containment venting, which is defined to be a loss of the primary containment barrier.

• Atmosphere samples from a minimum of two locations inside the primary containment and one location in the suppression chamber are sequentially monitored for hydrogen and oxygen percentage levels by each of two redundant analyzer systems: The analyzers are single range (0 to 30% hydrogen and Oto 30% oxygen). Two redundant (divisional) recorders are provided in the Main Control Room CMS 02/H2R 1 (H13 P827) and CMS 02/H2R 2 (H13 P811).

Hydrogen and oxygen concentrations can also be displayed on the plant computers (Ref. 9-12)

• Any unplanned rise in PC pressure in the Cold Shutdown or Refueling mode indicates Containment Closure cannot be assured and the primary containment cannot be relied upon as a barrier to fission product release.

• RB (Reactor Building) area radiation monitors should provide indication of increased release that may be indicative of a challenge to Containment Closure. The Maximum Safe Operating levels are indicative of problems in the secondary containment that are spreading. The locations into which the primary system discharge is of concern correspond to the areas addressed in Table 24 of the EOP flowcharts (ref.13).

• This IC addresses the inability to restore and maintain reactor vessel level above the top of active fuel with containment challenged. This condition represents actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA PAG exposure levels offsite for more than the immediate site area.

Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable .

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 44 of 190 With CONTAINMENT CLOSURE not established, there is a high potential for a direct and unmonitored release of radioactivity to the environment. If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, then declaration of a General Emergency is not required.

The existence of an explosive mixture means, at a minimum, that the containment atmospheric hydrogen concentration is sufficient to support a hydrogen burn (i.e., at the lower deflagration limit). A hydrogen burn will raise containment pressure and could result in collateral equipment damage leading to a loss of containment integrity. It therefore represents a challenge to Containment integrity.

In the early stages of a core uncovery event, it is unlikely that hydrogen buildup due to a core uncovery could result in an explosive gas mixture in containment. If all installed hydrogen gas monitors are out-of-service during an event leading to fuel cladding damage, it may not be possible to obtain a containment hydrogen gas concentration reading as ambient conditions within the containment will preclude personnel access. During periods when installed containment hydrogen gas monitors are out-of-service, operators may use the other listed indications to assess whether or not containment is challenged.

The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operations at Commercial Nuclear Power Plants in the United States; and NU MARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

CGS Basis Reference(s):

1. SOP-EDR-OPS Equipment Drain System Operation

2. SOP-FDR-OPS Floor Drain System Operation

3. SOP-RHR-SDC RHR Shutdown Cooling

4. PPM 5.2.1 Primary Containment Control

5. PPM 5.3.1 Secondary Containment Control

6. BWROG EPG/SAG Revision 2, Sections PC/G

7. PPM 5.7.1 RPVand Primary Containment Flooding SAG, Table 19

8. PPM 5.2.1 Primary Containment Control

9. FSAR Section 7.5.1.5.4

10. PPM 5.0.10 Flowchart Training Manual

11. PPM 4.814.J1 814.J1 Annunciator Panel Alarms, 2-2

12. PPM 4.814.J2 814.J2 Annunciator Panel Alarms, 2-2 13 . PPM 5.3.1 Secondary Containment Control

14. NEI 99-01 CG1 Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency- Technical Bases Page: 45 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 1 - RPV Level Initiating Condition: Loss of RPV inventory affecting core decay heat removal capability EAL:

CS1 .2 Site Area Emergency RPV level cannot be monitored for GE 30 min. (Note 1)

AND Core uncovery is indicated by any of the following:

• UNPLANNED wetwell level rise GT 2 inches (PPM 5.2.1 entry condition)

• VALID indication of RB room flooding as identified by high level alarms (PPM 5.3.1 Table 25)

• Observation of UNISOLABLE RCS leakage outside primary containment of sufficient magnitude to indicate core uncovery Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

Basis:

In this EAL, all RPV level indication is unavailable and the RPV inventory loss must be detected by the leakage indications provided. Level increases must be evaluated against other potential sources of leakage such as cooling water sources inside the drywell to ensure they are indicative of RPV leakage.

Reactor Building equipment or floor drain sump level rise may be indicative of RPV inventory losses external to the primary containment from systems connected to the RPV (ref. 1, 2). With RHR System operating in the Shutdown Cooling mode, an unexplained rise in wetwell level could be indicative of RHR valve misalignment or leakage (ref. 3). If the make-up rate to the RPV unexplainably rises above the pre-established rate, a loss of RPV inventory may be occurring even if the source of the leakage cannot be immediately identified.

An UNPLANNED wetwell level increase to GT 2 inches Qr a VALID RB room high level alarm indicates a significant loss of RCS that could lead to core uncovery if not isolated (ref. 4, 5).

Visual observation of significant leakage from systems connected to the RCS in areas outside the primary containment that cannot be isolated could be indicative of a loss of RPV inventory sufficient to lead to core uncovery.

This IC addresses a significant and prolonged loss of reactor vessel/RCS inventory control and makeup capability leading to IMMINENT fuel damage. The lost inventory may be due to a RCS component failure, a loss of configuration control or prolonged boiling of reactor coolant. These conditions entail major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration .

Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency~ Technical Bases Page: 46 of 190 Following an extended loss of core decay heat removal and inventory makeup, decay heat will cause reactor coolant boiling and a further reduction in reactor vessel level. If RPV level cannot be restored, fuel damage is probable.

The 30-minute criterion is tied to a readily recognizable event start time (i.e., the total loss of ability to monitor level), and allows sufficient time to monitor, assess and correlate reactor and plant conditions to determine if core uncovery has actually occurred (i.e., to account for various accident progression and instrumentation uncertainties). It also allows sufficient time for performance of actions to terminate leakage, recover inventory control/makeup equipment and/or restore level monitoring.

The inability to monitor RPV level may be caused by instrumentation and/or power failures, or water level dropping below the range of available instrumentation. If water level cannot be monitored, operators may determine that an inventory loss is occurring by observing changes in sump and/or tank levels. Sump and/or tank level changes must be evaluated against other potential sources of water flow to ensure they are indicative of leakage from the RCS.

This EAL addresses concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal; SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues; NUREG-1449, Shutdown and Low-Power Operations at Commercial Nuclear Power Plants in the United States; and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management.

Escalation of the emergency classification level would be via IC CG1 or RG1 CGS Basis Reference(s):

• 1.

2.

3.

4.

5.

SOP-EDR-OPS Equipment Drain System Operation SOP-FDR-OPS Floor Drain System Operation SOP-RHR-SDC RHR Shutdown Cooling PPM 5.2.1 Primary Containment Control PPM 5.3.1 Secondary Containment Control

6. NEI 99-01 CS1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 47 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 2 - Loss of Emergency AC Power Initiating Condition: Loss of fill offsite and fill onsite AC power to emergency buses for 15 minutes or longer EAL:

CA2.1 Alert Loss of all offsite and all onsite AC power capability to emergency buses SM-7 and SM-8 for GE 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

4 5 def Basis:

This cold condition EAL is equivalent to the hot condition EAL MS1 .1.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

When in the cold shutdown, Refuel, or defueled mode, this condition is not classified as a Site Area Emergency because of the increased time available to restore an emergency bus to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition represents an actual or potential substantial degradation of the level of safety of the plant.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via IC CS1 or RS1.

CGS Basis Reference(s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. ABN-ELEC-LOOP Loss Of All Off-Site Electrical Power

7. SOP-ELECT-BACKFEED

8. NEI 99-01 CA2

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 48 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 2 - Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL:

CU2.1 Unusual Event AC power capability, Table 2, to emergency buses SM-7 and SM-8 reduced to a single power source for GE 15 min. (Note 1)

AND Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 2 AC Power Sources Offsite Startup Transformer TR-S Backup Transformer TR-B Backfeed 500 KV power through Main Transformers (if already aligned in modes 4, 5, def only)

Onsite

• DG1

• DG2

• Main Generator via TR-N1/N2 Mode Applicability:

4 5 def Basis:

Table 2 provides the list of AC power sources available to power emergency buses (ref. 1, 2).

Station Startup 230KV power comes from the Ashe substation through Startup transformer TR-S. The Startup transformer usually supplies station auxiliary loads when the main generator is not available.

Station Backup 115KV power from the Benton Substation feeder can be supplied to emergency buses SM-7 and SM-8 (ref. 3, 4) .

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency -Technical Bases Page: 49 of 190 Credit is not taken in this EAL for SM-4/DG3 crosstie capability because establishing the crosstie to SM-7 or SM-8 is assumed to require more than 15 minutes (5). SM-4 is not a site specific emergency AC buss source since SM-4 does not provide core cooling or containment cooling.

It is possible to remove Startup power from service and continue to supply the plant during shutdown conditions by backfeeding 500 KV power from Ashe Substation through the Main Transformers, the Normal Transformers and associated "N" breakers. This involves disconnecting the Main Generator from the Isolated Phase conductors (25 KV system) and overriding various interlocks. This action would take significantly longer than 15 minutes; therefore, backfeed must be in service to credit this source (r~n. .

The second threshold statement in this EAL does not describe a separate condition; it is clarifying the first threshold statement.

This cold condition EAL is equivalent to the hot condition EAL MA 1.1.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

When in the cold shutdown, Refuel, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures and pressures in various plant systems. Thus, when in these modes, this condition is considered to be a potential

• degradation of the level of safety of the plant.

An "AC power source" is a source recognized in AOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

• A loss of all offsite power with a concurrent failure of one division of emergency power sources (e.g., onsite diesel generators).

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

• A loss of emergency power sources (e.g., onsite diesel generators) with a single division of emergency buses being back-fed from an offsite power source.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

The subsequent loss of the remaining single power source would escalate the event to an Alert in accordance with IC CA2.

CGS Basis Reference(s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. ABN-ELEC-LOOP Loss Of All Off-Site Electrical Power

7. SOP-ELECT-BACKFEED

• 8. NEI 99-01 CU2 Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 50 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 3- RCS Temperature Initiating Condition: Inability to maintain the plant in cold shutdown EAL:

CA3.1 Alert UNPLANNED increase in RCS temperature to GT 200°F for GT Table 7 duration (Note 1)

OR UNPLANNED RPV pressure increase GT 10 psig Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceed e,d orw1*11 l"k I be excee ded 1 e1y Table 7 RCS Heat-up Duration Thresholds CONTAINMENT CLOSURE RCS Status Heat-up Duration Status Intact N/A 60 min.*

established 20 min.*

Not intact not established O min.

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable.

Mode Applicability:

Basis:

200°F is the Technical Specification cold shutdown temperature limit (ref. 1).

10 psi is one-half of the 20 psi minor division on the Wide Range RPV pressure instrument, RFW-Pl-605, on Main Control Room Panel H13- P603 (ref. 2). This instrument has a range of Oto 1200 psig. This RPV pressure indication is also displayed on plant computer point 8016 (ref. 3).

Recirculation suction temperature, RRC TR 650 pt 1(2), is the primary temperature measurement

• instrument when RPV pressure is less than 100 psig and the associated RRC pump is operating.

Monitoring of the RWCU bottom head drain temperature element, RWCU TE 21, as read on RWCU Tl 607 pt 5 (H13 P602) or MS TR 6 pt 316 (RB 522) is acceptable only if a RRC pump is operating for forced flow and RWCU flow of greater than 50 gpm exists. (ref. 4)

With flow through the RHR Heat Exchanger, the inlet temperature (TDAS pt. X045) is indicative of RRC system temperature. If adequate core flow cannot be provided, RPV metal temperature can be

• monitored on MS TR 6. (ref. 5)

Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 51 of 190 The RCS should be considered intact when the RCS pressure boundary is in its normal condition for the cold shutdown mode of operation (e.g., no freeze seals or nozzle dams).

This IC addresses conditions involving a loss of decay heat removal capability or an addition of heat to the RCS in excess of that which can currently be removed. Either condition represents an actual or potential substantial degradation of the level of safety of the plant.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

The RCS Heat-up Duration Thresholds table addresses an increase in RCS temperature when CONTAINMENT CLOSURE is established but the RCS is not intact.. The 20-minute criterion was included to allow time for operator action to address the temperature increase.

The RCS Heat-up Duration Thresholds table also addresses an increase in RCS temperature with the RCS intact. The status of CONTAINMENT CLOSURE is not crucial in this condition since the intact RCS is providing a high pressure barrier to a fission product release. The 60-minute time frame should allow sufficient time to address the temperature increase without a substantial degradation in plant safety.

Finally, in the case where there is an increase in RCS temperature, the RCS is not intact , and CONTAINMENT CLOSURE is not established, no heat-up duration is allowed (i.e., 0 minutes). This is because 1) the evaporated reactor coolant may be released directly into the Containment atmosphere and subsequently to the environment, and 2) there is reduced reactor coolant inventory above the top

• of irradiated fuel.

The RCS pressure increase threshold provides a pressure-based indication of RCS heat-up in the absence of RCS temperature monitoring capability.

Escalation of the emergency classification level would be via IC CS1 or RS1.

CGS Basis Reference(s):

1. Technical Specifications Table 1.1-1

2. Instrument Master Datasheet for EPN RFW-Pl-605

3. PPM 10.27.36 Reactor Pressure High Alarm - CC

4. OSP-RCS-C102 RPV Non-Critical Cooldown Surveillance

5. SOP-RHR-SDC RHR Shutdown Cooling

6. NEI 99-01 CA3

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 52 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 3- RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.1 Unusual Event UNPLANNED increase in RCS temperature to GT 200°F Mode Applicability:

Basis:

In the absence of reliable RCS temperature indication caused by a loss of decay heat removal capability, classification should be based on EAL CU3.2 should RCS level indication be subsequently lost.

The Technical Specification cold shutdown temperature limit is 200°F (ref. 1).

This IC addresses an UNPLANNED increase in RCS temperature above the Technical Specification cold shutdown temperature limit and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3.

A momentary UNPLANNED excursion above the Technical Specification cold shutdown temperature limit when the heat removal function is available does not warrant a classification.

This EAL involves a loss of decay heat removal capability, or an addition of heat to the RCS in excess of that which can currently be removed, such that reactor coolant temperature cannot be maintained below the cold shutdown temperature limit specified in Technical Specifications. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of Power Operations.

During an outage, the level in the reactor vessel will normally be maintained at or above the reactor vessel flange. Refuel evolutions that lower water level below the reactor vessel flange are carefully planned and controlled. A loss of forced decay heat removal at reduced inventory may result in a rapid increase in reactor coolant temperature depending on the time after shutdown.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

CGS Basis Reference(s):

1. Technical Specifications Table 1.1-1

2. NEI 99-0~ CU3

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 53 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 3 - RCS Temperature Initiating Condition: UNPLANNED increase in RCS temperature EAL:

CU3.2 Unusual Event Loss of all RCS temperature and RPV water level indication for GE 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

Basis:

Recirculation suction temperature, RRC-TR-650 pt 1(2), is the primary temperature measurement instrument when RPV pressure is less than 100 psig and the associated RRC pump is operating.

Monitoring of the RWCU bottom head drain temperature element, RWCU-TE-21, as read on RWCU-Tl-607 pt 5 (H13 P602) or MS-TR-6 pt 316 (RB 522) is acceptable only if a RRC pump is operating for

• forced flow and RWCU flow of greater than 50 gpm exists. (ref. 4)

With flow through the RHR Heat Exchanger, the inlet temperature (TDAS pt. X045) is indicative of RRC system temperature. If adequate core flow cannot be provided, RPV metal temperature can be monitored on MS-TR-6. (ref. 5)

This EAL addresses the inability to determine RCS temperature and RPV level, and represents a potential degradation of the level of safety of the plant. If the RCS is not intact and CONTAINMENT CLOSURE is not established during this event, the Emergency Director should also refer to IC CA3.

This EAL reflects a condition where there has been a significant loss of instrumentation capability necessary to monitor RCS conditions and operators would be unable to monitor key parameters necessary to assure core decay heat removal. During this condition, there is no immediate threat of fuel damage because the core decay heat load has been reduced since the cessation of Power Operations.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation to Alert would be via IC CA1 based on an inventory loss or IC CA3 based on exceeding plant configuration-specific time criteria.

CGS Basis Reference(s):

1. FSAR Table 7.5-1

2. FSAR Figure 7.7-1

3. FSAR Section 7.6.1.3

4. OSP-RCS-C102 RPV Non-Critical Cooldown Surveillance

5. SOP-RHR-SDC RHR Shutdown Cooling

6. NEI 99-01 CU3

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 54 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 4 - Loss of Vital DC Power Initiating Condition: Loss of vital DC power for 15 minutes or longer EAL:

CU4.1 Unusual Event Indicated voltage LT 108 voe on required 125 voe buses DP-S1-1 and DP-S1-2 for GE 15 min.

(Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

Basis:

The 125 VDC Class 1E DC power system consists of three electrically independent and separate distribution systems (S1-1, S1-2, and S1-HPCS). S1-HPCS is not included in this EAL. Each DC distribution system has a battery and a battery charger that are normally connected to the bus such that these two sources of power are operating in parallel. The charger is normally supplying system

• electrical loads with the battery on a float charge. Each battery has the necessary amp-hour discharge capacity to sustain system loads for a minimum of two hours. This capacity is specifically for a loss of power to the charger coincident with a design basis accident. The batteries have capacity to carry design load at 60°F without decreasing battery voltage below 1.81 volts/cell (or 108 VDC, ref. 1) with loss of output from the battery chargers during the specified period. Battery capacity is sufficient to provide starting currents while operating at full load. (ref. 2)

This EAL is the cold condition equivalent of the hot condition loss of DC power EAL MS2.1.

This IC addresses a loss of essential DC power which compromises the ability to monitor and control operable SAFETY SYSTEMS when the plant is in the cold shutdown or Refuel mode. In these modes, the core decay heat load has been significantly reduced, and coolant system temperatures and pressures are lower; these conditions increase the time available to restore a vital DC bus to service.

Thus, this condition is considered to be a potential degradation of the level of safety of the plant.

As used in this EAL, "required" means the essential DC buses necessary to support operation of the in-service, or operable, train or trains of SAFETY SYSTEM equipment. For example, if Division I is out-of-service (inoperable) for scheduled outage maintenance work and Division II is in-service (operable),

then a loss of essential DC power affecting Division II would require the declaration of an Unusual Event. A loss of essential DC power to Division. I would not warrant an emergency classification.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Depending upon the event, escalation of the emergency classification level would be via IC CA 1 or CA3, or an IC in Recognition Category R.

CGS Basis Reference(s): ,

1. Calculation No. 2.05.01 Battery Sizing, Voltage Drop, and Charger Studies for Div. 1 & 2 Systems

2. FSAR Section 8.3.2

3. NEI 99-01 CU4 Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 55 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 5 - Loss of Communications Initiating Condition: Loss of all onsite or offsite communications capabilities EAL:

CU5.1 Unusual Event (1) Loss of all Table 4 onsite communication methods OR (2) Loss of all Table 4 ORO communication methods OR (3) Loss of all Table 4 NRC communication methods Table 4 Communication Methods System Onsite ORO NRC Plant Public Address (PA) System X

• Plant Telephone System Plant Radio System Operations and Security Channels Offsite calling capability from the Control X

X X

X X Room via direct telephone Long distance calling capability on the X X commercial phone system Mode Applicability:

4 5 def Basis:

Onsite and offsite (ORO and NRC) communications include one or more of the systems listed in Table 4 (ref. 1, 2).

Public Address (PA) System The public address system provides a way of contacting personnel in the various buildings of the plant and locations of the site that might be inaccessible using other means of communication. The building-wide alarm system alerts (via the public address system speakers) operating personnel to fire hazards and other trouble conditions for which plant management finds it necessary to alert plant personnel.

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency-Technical Bases Page: 56 of 190 Plant Telephone System This system consists of interconnections to the public telephone network (and trunks to the PBX) with individual direct lines that provide inward and outward dialing access to most plant locations.

Plant Radio System Operations and Security Channels The radio communications system is used for communications with personnel involved in maintenance and security in and around the plant complex by means of hand-held portable radio units, mol;>ile radio units, and paging receivers. The telephone link to BPA provides a direct communication link to the BPA Dittmer Control Center. The radio communications system provides a communications link for security and emergency communications to local law enforcement agencies and emergency control centers.

Offsite calling capability from the Control Room via direct telephone and fax lines This communications method includes following dedicated Rhone networks that are available for emergency communications ,in addition to the normal Energy Northwest phone network:

• Energy Northwest Emergency Center Network

• Response Agency Network

• NRC Emergency Notification System Various locations such as the Control Room, Technical Support Center, Emergency Operations Facility, Joint Information Center, Department of Energy-RL, Washington State Emergency Operations Center,

• Oregon State Emergency Coordination Center and the Benton and Franklin County Emergency Operations Centers have facsimile transceivers. The facsimile transceivers enable the transmission and receipt of printed material. The facsimile system which connects the Energy Northwest emergency centers with the county and state emergency centers uses dedicated phone lines.

Long distance calling capability on the commercial phone system The Energy Northwest Richland phone system is a computer based, software controlled telephone exchange (Computerized Branch Exchange). It is equipped with redundant computerized processor units and is served by an uninterruptible power supply. The direct-dial private telephone system provides communication between the Energy Northwest facilities. The phone system is arranged such that plant telephones can reach other Energy Northwest facilities by direct-dialing and without the need of an operator.

This EAL is the cold condition equivalent of the hot condition EAL MU7.1.

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct

, challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.

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

The first EAL condition addresses a total loss of the communications methods used in support of routine plant operations ..

The second EAL condition addresses a total loss of the communications methods used to notify all OROs of an emergency declaration. The OROs referred to here are Washington Stare, Benton County, Franklin County and DOE RL.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A J Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 57 of 190 The third EAL addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

CGS Basis Reference(s):

1. Emergency Plan Section 6.6

2. FSAR Section 9.5.2

3. NEI 99-01 CU5

• Attachment 6.1, EAL Technical Bases

I Use Category:

Number: 13.1.1A REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 58 of 190 Category: C - Cold Shutdown / Refuel System Malfunction Subcategory: 6 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:

CA6.1 Alert The occurrence of any Table 8 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to a second train of a SAFETY SYSTEM needed for the current operating mode OR Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating mode

• Note 9:

(Notes 9, 10)

If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Table 8 Hazardous Events

• Seismic event

• Internal or external FLOODING event

• High winds

• Tornado strike

• FIRE

• EXPLOSION

• Volcanic ash fallout

• Other events with similar hazard characteristics as determined by the Shift Manager

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 59 of 190 Basis:

This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

• VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

The significance of a seismic event is 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).

Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 100 mph (ref. 3).

Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Areas in the fire response procedure (ref. 4).

The potential for volcanic eruption exists in the Pacific Northwest. Heavy ash fall, such as that experienced at certain lo~ations following the eruption of Mt. St. Helens in 1980, could affect operation of plant equipment if precautionary measures are not taken. The design basis ash fall is projected for a twenty hour duration. (ref. 5)

Table 5 provides a list of CGS safety system areas (ref. 6).

Escalation of the emergency classification level would be via IC CS1 or RS1 .

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 60 of 190 CGS Basis Reference(s):

1. FSAR Section 3.7 Seismic Design

2. FSAR Section 3.4.1 Flood Protection

3. CGS Calculation CALC CE-02-93-16 Evaluate PMR/BDC 98-0131-0A change from 5 min. to 15 min. averaging of 33 ft. elev. met tower wind speeds for UE and Alert declarations

4. ABN-FIRE Attachment 13.2, Fire Areas

5. ABN-ASH Ash Fall

6. FSAR Table 3.2-1 Equipment Classification

7. NEI 99-01 CA6

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 61 of 190 Category E - Independent Spent Fuel Storage Installation (ISFSI)

EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold)

An independent spent fuel storage installation (ISFSI) is a complex that is designed and constructed 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 a significant environmental effect resulting from an accident involving the dry storage of spent nuclear fuel.

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 a classifiable event under Security category EAL HA 1.1 .

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 62 of 190 Category: E- ISFSI Sub-category: None Initiating Condition: Damage to a loaded cask CONFINEMENT BOUNDARY EAL:

EU1.1 Unusual Event Damage to a loaded canister (MPC) CONFINEMENT BOUNDARY as indicated by measured dose rates on a loaded overpack GT EITHER:

• 20 mrem/hr (gamma + neutron) on the top of the overpack

• 100 mrem/hr (gamma + neutron) on the side of the overpack, excluding inlet and outlet ducts Mode Applicability:

Storage Operations Basis:

The Independent Spent Fuel Storage Installation utilizes the HOLTEC International (HOLTEC) HI-STORM 100 Spent Fuel Dry Storage (SFDS) system. HI-STORM overpack or storage overpack means the cask that receives and contains the sealed multi-purpose canisters containing spent nuclear fuel. It provides the gamma and neutron shielding, ventilation passages, missile protection, and protection against natural phenomena and accidents for the MPC. (ref. 1, 2)

The EAL threshold values represent two-times the limits specified in the ISFSI Certificate of Compliance Technical Specification Section 3.2, Radiation Protection Program (ref. 2).

CGS has casks loaded to various amendments to the Certificate of Compliance (COC) Technical Specifications. The numbers above reflect the most limiting Technical Specification (TS) values (Amendment 1).

This IC addresses an event that results in damage to the CONFINEMENT BOUNDARY of a storage cask containing spent fuel. It applies to irradiated fuel that is licensed for dry storage beginning at the point that the loaded storage cask is sealed. The issues of concern are the creation of a potential or actual release path to the environment, degradation of one or more fuel assemblies due to environmental factors, and configuration changes which could cause challenges in removing the cask or fuel from storage.

The existence of "damage" is determined by radiological survey. The technical specification multiple of "2 times", which is also used in Recognition Category RIC RU1, is used here to distinguish between non-emergency and emergency conditions. The emphasis for this classification is the degradation in the level of safety of the spent fuel cask and not the magnitude of the associated dose or dose rate. It is recognized that in the case of extreme damage to a loaded cask, the fact that the "on-contact" dose rate limit is exceeded may be determined based on measurement of a dose rate at some distance from the cask.

Security-related events for ISFSls are covered under ICs HU 1 and HA 1 .

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major. Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 63 of 190 CGS Basis Reference(s):

1. ABN-ISFSI, ISFSI Abnormal Conditions

2. ISFSI Certificate of Compliance No. 1014 Amendment 1, Appendix A, Technical Specifications for the HI-STORM 100 Cask System, Section 3.2 Radiation Protection Program

3. NEI 99-01 E-HU1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A iuseC~ego~:REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency -Technical Bases Page: 64 of 190 Fission Product Barrier Matrix and Bases Introduction Table F-1 lists the threshold conditions that define the Loss and Potential Loss of the three fission product barriers (Fuel Clad, Reactor Coolant System, and Containment). The table is structured so that each of the three barriers occupies adjacent columns. Each fission product barrier column is further divided into two columns; one for Loss thresholds and one for Potential 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. RPV Water Level B. RCS Leak Rate c.- PC Conditions D. PC Radiation / RCS Activity E. PC Integrity or Bypass F. Emergency Director Judgment Each catego~ 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 one or more fission product barrier thresholds appear. If NEI 99-01 does not define a threshold for a barrier Loss/Potential Loss, the word "None" is entered in the cell.

Thresholds are assigned sequential numbers within each Loss and Potential Loss column beginning with number one. In this manner, a threshold can be identified by its catego~ title and number. For example, the first Fuel Clad barrier Loss in Catego~ A would be assigned "FC Loss A.1," the third Containment barrier Potential Loss would be assigned "PC P-Loss B.3," etc.

If a cell in Table F-1 contains more than one numbered threshold, each of the numbered thresholds, if exceeded, signifies a Loss or Potential Loss of the barrier. It is not necessa~ to exceed all of the thresholds in a catego~ before declaring a barrier Loss/Potential Loss.

Subdivision of Table F-1 by catego~ facilitates association of plant conditions to the applicable fission product barrier Loss and Potential Loss thresholds. This structure promotes a 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, the EAL-user first scans down the catego~ column of Table F-1, locates the likely catego~ and then reads across the fission product barrier Loss and Potential Loss thresholds in that catego~ to determine if a threshold has been exceeded. If a threshold has not been exceeded, the EAL-user proceeds to the next likely catego~ and continues review of the thresholds in the new catego~.

If the EAL-user determines that any threshold has been exceeded, by definition, the barrier is lost 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 three fission product barriers to determine if other barrier thresholds in the catego~ are lost or potentially lost. For example, if containment radiation is sufficiently high, a Loss of the Fuel Clad and RCS barriers and a Potential Loss of the prima~ containment barrier can occur. Barrier Losses and Potential Losses are then applied to the algorithms given in EALs FG1 .1, FS1 .1, and FA 1.1 to determine the appropriate emergency classification.

The Fuel Clad barrier threshold bases appear first, followed by the RCS barrier and finally the Containment barrier threshold bases. In each barrier, the bases are given according catego~ Loss followed by catego~ Potential Loss beginning with Catego~ A, then B, ... , F. Blank sections of the table do not have basis pages.

Attachment 6.1, EAL Technical Bases

• Number: 13.1.1A

• I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Er:nergency - Technical Bases Page: 65 of 190 Table F-1 Fission Product Barrier Threshold Matrix FC - Fuel Clad Barrier RCS - Reactor Coolant System Barrier PC - Containment Barrier Category Loss Potential Loss Loss Potential Loss Loss Potential Loss A RPV level cannot be restored and RPV level cannot be restored and SAG entry required maintained GT -161 in. maintained GT -161 in. None None SAG entry required RPVWater or cannot be determined or cannot be determined Level UNISOLABLE break in fillY of the following:

Main steam lines UNISOLABLE primary system leakage UNISOLABLE primary system leakage that results in exceeding

. RCIC steam Line that results in exceeding EITHER:

EITHER:

B . RWCU RB area temperature alarm level (PPM 5.3.1 Table 23)

RB area maximum safe operating None RCS Leak Rate None None

. Feedwater OR RB area radiation alarm level temperature (PPM 5.3.1 Table 23)

OR OR RB area maximum safe operating (PPM 5.3.1 Table 24) radiation (PPM 5.3.1 Table 24)

Emergency RPV Depressurization is required PC pressure GT 45 psig UNPLANNED rapid drop in PC OR C PC pressure GT 1.68 psig pressure following PC pressure rise Explosive mixture exists inside PC (H 2 None None None OR GE 6% and o,GE 5%)

PC due to RCS leakage PC pressure response !1Q1_consistent OR Conditions WW temperature and RPV pressure with LOCA conditions cannot be maintained below the HCTL Containment Radiation Monitor D CMS-RIS-27E or CMS-RIS-27F Containment Radiation Monitor CMS- Containment Radiation Monitor CMS-reading GT 3,600 R/hr PC Rad/ None RIS-27E or CMS-RIS-27F reading None None RIS-27E or CMS-RIS-27F reading OR RCS GT 70 R/hr GT 14,000 R/hr Primary coolant activity GT 300 Activity

µCi/qm dose equivalent 1-131 UNISOLABLE direct downstream E pathway to the environment exists None None None None after PC isolation signal None PC Integrity OR or Bypass Intentional PC venting per EOPs F 8rJ1. condition in the opinion of the 8rJ1. condition in the opinion of the 8rJ1. condition in the opinion of the 8rJ1. condition in the opinion of the 8rJ1. condition in the opinion of the 8rJ1. condition in the opinion of the Emergency Director that indicates Emergency Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates potential loss of the Containment Director loss of the fuel clad barrier potential loss of the fuel clad barrier loss of the RCS barrier potential loss of the RCS barrier loss of the Containment barrier barrier Judgment Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: o.33 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 66 of 190 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss of any two barriers and loss or potential loss of third barrier EAL:

FG1.1 General Emergency Loss of any two barriers AND Loss or potential loss of third barrier (Table F-1)

Mode Applicability:

1 2 3 Basis:

Fuel Clad, RCS 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 General Emergency is therefore appropriate for any combination of the following conditions:

• Loss of Fuel Clad, RCS and Containment barriers

• Loss of Fuel Clad and RCS barriers with potential loss of Containment barrier

• Loss of RCS and Containment barriers with potential loss of Fuel Clad barrier

• Loss of Fuel Clad and Containment barriers with potential loss of RCS barrier CGS Basis Reference(s):

1. NEI 99-01 FG1

• Attachment 6.1, EAL Technical Bases

I IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 67 of 190 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Loss or potential loss of any two barriers EAL:

FS1 .1 Site Area Emergency Loss or potential loss of any two barriers (Table F-1)

Mode Applicability:

1 2 3 Basis:

Fuel Clad, RCS 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 Area Emergency 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. For example, the existence of Fuel Clad and RCS Barrier loss thresholds in addition to offsite dose assessments would require continual assessments of radioactive inventory and Containment integrity in anticipation of reaching a General Emergency classification. Alternatively, if both Fuel Clad and RCS potential loss thresholds existed, the Emergency Director would have greater assurance that escalation to a General Emergency is less imminent.

CGS Basis Reference(s):

1. NEI 99-01 FS1

\

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 68 of 190 Category: Fission Product Barrier Degradation Subcategory: N/A Initiating Condition: Any loss or any potential loss of either Fuel Clad or RCS EAL:

FA1.1 Alert Any loss or any potential loss of EITHER Fuel Clad or RCS barrier (Table F-1)

Mode Applicability:

1 2 3 Basis:

Fuel Clad, RCS and Containment comprise the fission product barriers.

Table F-1 (Attachment 6.2) lists the fission product barrier thresholds, bases and references.

At the Alert classification level, Fuel Clad and RCS barriers are weighted more heavily than the Containment barrier. Unlike the Containment barrier, loss or-potential loss of either the Fuel Clad or RCS barrier may result in the relocation of radioactive materials or degradation of core cooling capability. Note that the loss or potential loss of Containment barrier in combination with loss or potential loss of either Fuel Clad or RCS barrier results in declaration of a Site Area Emergency under

• EAL FS1.1 CGS Basis Reference(s):

1. NEI 99-01 FA1

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 69 of 190 Barr:ier: Fuel Clad Category: A. RPV Level Degradation Threat: Loss Threshold:

I SAG entry required Basis:

EOP flowcharts provide instructions to assure adequate core cooling by restoring and maintaining RPV water level above prescribed limits, operate sufficient RPV injection sources to assure adequate core cooling, and assess the possibility of core damage when RPV level cannot be determined. The Fuel Clad Loss threshold conditions are the EOP flowchart conditions that signal a loss of adequate core cooling and a requirement to exit all EOPs and enter the SAGs (ref. 1-6).

This threshold is also a Loss of the RCS barrier (RCS Loss A) and a Potential Loss of the Containment barrier (PC P-Loss A), and therefore represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification.

• The Loss threshold represents the EOP requirement for entry to the Severe Accident Guidelines (SAGs) .

CGS Basis Reference(s):

1.

2.

3.

PPM 5.1.1 RPV Control PPM 5.1.2 RPV Control - A TWS Calculation NE-02-03-06 Attachment 10 RPV Variables

4. PPM 5.0.1 O Flowchart Training Manual

5. PPM 5.1.4 RPV Flooding

6. PPM 5.1.6 RPV Flooding - ATWS

7. NEI 99-01 RPV Water Level Fuel Clad Loss 2.A

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency- Technical Bases Page: 70 of 190 Barrier: Fuel Clad Category: A. RPV Level Degradation Threat: Potential Loss Threshold:

RPV level cannot be restored and maintained GT -161 in. or cannot be determined Basis:

An RPV water level instrument reading of-161 in. indicates RPV level is at the top of active fuel (TAF)

(ref. 1, 2). When RPV level is at or above the TAF, the core is completely submerged. Core submergence is the most desirable means of core cooling. When RPV level is below TAF, the uncovered portion of the core must be cooled by less reliable means (i.e., steam cooling or spray cooling). If core uncovery is threatened, the EOPs specify alternate, more extreme, RPV water level control measures in order to restore and maintain adequate core cooling. Since core uncovery begins if RPV level drops to TAF, the level is indicative of a challenge to core cooling and the Fuel Clad barrier.

When RPV level cannot be determined, EOPs require RPV flooding strategies. RPV water level

• indication provides the primary means of knowing if adequate core cooling is being maintained. When all means of determining RPV level are unavailable, the fuel clad barrier is threatened and reliance on alternate means of assuring adequate core cooling must be attempted. The instructions in PPM 5.1.4 and PPM 5.1.6 specify these means, which include emergency depressurization of the RPV and injection into the RPV at a rate needed to flood to the elevation of the main steam lines or hold RPV pressure above the Minimum Steam Cooling Pressure (in A TWS events) (ref. 3, 4). If RPV level cannot be determined with respect to the top of active fuel, a potential loss of the fuel clad barrier exists.

This water level corresponds to the top of the active fuel and is used in the EOPs to indicate a challenge to core cooling.

The RPV water level threshold is the same as RCS barrier Loss RPV Water Level threshold .A. Thus, this threshold indicates a Potential Loss of the Fuel Clad barrier and a Loss of the RCS barrier that appropriately escalates the emergency classification level to a Site Area Emergency.

This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this Fuel Clad barrier Potential Loss is met only after either:

1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or

2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory.

Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 71 of 190 The term "cannot be restored and maintained above" means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel; but does not permit extended operation below the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.

In high-power ATWS/failure to scram events, EOPs may direct the operator to deliberately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs MAS or MS6 will dictate the need for emergency classification.

Since the loss of ability to determine if adequate core cooling is being provided presents a significant challenge to the fuel clad barrier, a potential loss of the fuel clad barrier is specified.

CGS Basis Reference(s):

1. Calculation NE-02-03-05 Attachment 3 Note 8

2. PPM 5.1.1 RPV Control

3. PPM 5.1.4 RPV Flooding

• 4.

5.

6.

PPM 5.1.6 RPV Flooding -ATWS PPM 5.1.2 RPV Control - A TWS NEI 99-01 RPV Water Level Fuel Clad Potential Loss 2.A

• Attachment 6.1, EAL Technical Bases

• • Number: 13.1.1A

Title:

Classifying The Emergency - Technical Bases Barrier: Fuel Clad I Use Category: REFERENCE Major Rev: 033 Minor Rev:* NIA Page: 72 of 190 Category: D. PC Radiation/ RCS Activity Degradation Threat: Loss Threshold:

Containment Radiation Monitor CMS-RIS-27E or CMS-RIS-27F reading GT 3,600 R/hr Basis:

Four high range area radiation detectors (CMS-RE-27A, B, E and F) are installed to monitor the drywall. CMS-RE-27A and -278 are located in the bioshield wall at elevations 522' and 525', azimuth 60° and 297°, respectively. CMS-RE-27E and -27F are located inside containment at elevation 515',

azimuth 290° and 51.5°, respectively. The companion containment radiation monitors (CMS-RIS-27A, B, E and F) are located on RAD Boards 22 and 23 in the Main Control Room. (ref. 1)

The threshold value was calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with a concentration of 300 µCi/gm dose equivalent 1-131 (or approximately 5% clad failure) into the drywall atmosphere. Evaluation of detector location, geometry and anticipated response suggests CMS-RIS-27E or F will provide the desired response to a given radiation source in the drywall and are, therefore, identified as the preferred monitors for evaluating this Fuel Clad Loss threshold. (ref. 2)

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals 300 µCi/gm dose equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

The radiation monitor reading in this threshold is higher than that specified for RCS Barrier Loss threshold D since it indicates a loss of both the Fuel Clad Barrier and the RCS Barrier. Note that a combination of the two monitor readings appropriately escalates the emergency classification level to a Site Area Emergency.

There is no Potential Loss threshold associated with primary containment radiation.

CGS Basis Reference(s):

1. TM-2117 TSG - Core Thermal Engineer, Attachment 4.2

2. Calculation NE-02-94-57

2. NEI 99-01 Primary Containment Radiation Fuel Clad Loss 4.A

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 73 of 190 Barrier: Fuel Clad Category: D. PC Radiation/ RCS Activity Degradation Threat: Loss Threshold:

Primary coolant activity GT 300 µCi/gm dose equivalent 1-131 Basis:

This threshold indicates that RCS radioactivity concentration is greater than 300 µCi/gm Dose Equivalent 1-131. Reactor coolant activity above this level is greater than that expected for iodine spikes and corresponds to an approximate range of 2% to 5% fuel clad damage. Since this condition indicates that a significant amount of fuel clad damage has occurred, it represents a loss of the Fuel Clad Barrier.

It is recognized that sample collection and analysis of reactor coolant with highly elevated activity levels could require several hours to complete. Nonetheless, a sample-related threshold is included as a backup to other indications.

There is no Potential Loss threshold associated with RCS Activity.

• CGS Basis Reference(s):

1. NEI 99-01 RCS Activity Fuel Clad Loss 1.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 74 of 190 Barrier: Fuel Clad Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold:

Any condition in the opinion of the Emergency Director that indicates loss of the Fuel Clad barrier Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to 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 from portable instrumentation and consideration of offsite monitoring results .

Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is lost.

CGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Fuel Clad Loss 6.A

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 75 of 190 Barrier: Fuel Clad Category: F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:

Any condition in the opinion of the Emergency Director that indicates potential loss of the Fuel Clad barrier Basis:

' The Emergency Director judgment threshold addresses any other factors relevant to determining if the Fuel Clad barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to 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 from portable instrumentation and consideration of offsite monitoring results.

Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that are to be used by the Emergency Director in determining whether the Fuel Clad barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

CGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment Potential Fuel Clad Loss 6.A

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 76 of 190 Barrier: Reactor Coolant System Category: A. RPV Water Level Degradation Threat: Loss Threshold:

RPV level cannot be restored and maintained GT -161 in. or cannot be determined Basis:

An RPV water level instrument reading of-161 in. indicates level is at the top of active fuel (TAF)

(ref. 1, 2). TAF is significantly lower than the normal operating RPV level control band. To reach this level, RPV inventory loss would have previously required isolation of the RCS and PC barriers, and initiation of all ECCS. If RPV water level cannot be maintained above TAF, ECCS and other sources of RPV injection have been ineffective or incapable of reversing the decreasing level trend. The cause of the loss of RPV inventory is therefore assumed to be a LOCA. By definition, a LOCA event is a Loss of the RCS barrier.

When RPV water level cannot be determined, EOPs require RPV flooding strategies. RPV water level

• indication provides the primary means of knowing if adequate core cooling is being maintained. The instructions in PPM 5.1.4 and PPM 5.1.6 specify emergency depressurization of the RPV, which is defined to be a Loss of the RCS barrier (RCS Loss B threshold #2). (ref. 3, 4)

The conditions of this threshold are also a Potential Loss of the Fuel Clad barrier (FC P-Loss A). A Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier requires a Site Area Emergency classification.

This water level corresponds to the top of active fuel and is used in the EOPs to indicate challenge to core cooling.

The RPV water level threshold is the same as the Fuel Clad barrier RPV Water Level Potential Loss threshold. Thus, this threshold indicates a Loss of the RCS barrier and Potential Loss of the Fuel Clad barrier and that appropriately escalates the emergency classification level to a Site Area Emergency.

This threshold is considered to be exceeded when, as specified in the site-specific EOPs, RPV water cannot be restored and maintained above the specified level following depressurization of the RPV (either manually, automatically or by failure of the RCS barrier) or when procedural guidance or a lack of low pressure RPV injection sources preclude Emergency RPV depressurization. EOPs allow the operator a wide choice of RPV injection sources to consider when restoring RPV water level to within prescribed limits. EOPs also specify depressurization of the RPV in order to facilitate RPV water level control with low-pressure injection sources. In some events, elevated RPV pressure may prevent restoration of RPV water level until pressure drops below the shutoff heads of available injection sources. Therefore, this RCS barrier Loss is met only after either: 1) the RPV has been depressurized, or required emergency RPV depressurization has been attempted, giving the operator an opportunity to assess the capability of low-pressure injection sources to restore RPV water level or 2) no low pressure RPV injection systems are available, precluding RPV depressurization in an attempt to minimize loss of RPV inventory .

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 77 of 190 The term, "cannot be restored and maintained above," means the value of RPV water level is not able to be brought above the specified limit (top of active fuel). The determination requires an evaluation of system performance and availability in relation to the RPV water level value and trend. A threshold prescribing declaration when a threshold value cannot be restored and maintained above a specified limit does not require immediate action simply because the current value is below the top of active fuel, but does not permit extended operation beyond the limit; the threshold must be considered reached as soon as it is apparent that the top of active fuel cannot be attained.

In high-power ATWS/failure to scram events, EOPs may direct the operator to delib~rately lower RPV water level to the top of active fuel in order to reduce reactor power. RPV water level is then controlled between the top of active fuel and the Minimum Steam Cooling RPV Water Level (MSCRWL). Although such action is a challenge to core cooling and the Fuel Clad barrier, the immediate need to reduce reactor power is the higher priority. For such events, ICs MA6 or MS6 will dictate the need for emergency classification.

There is no RCS Potential Loss threshold associated with RPV Water Level.

CGS Basis Reference(s):

1. Calculation NE-02-03-05 Attachment 3 Note 8

2. PPM 5.1.1 RPV Control

3. PPM 5.1.4 RPV Flooding

4. PPM 5.1.6 RPV Flooding - ATWS

5. PPM 5.1.2 RPV Control - ATWS

6. NEI 99-01 RPV Water Level RCS Loss 2.A

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency- Technical Bases Page: 78 of 190 Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

UNISOLABLE break in any of the following:

• Main steam line

• RCIC steam line

• RWCU

• Feedwater Basis:

The conditions of this threshold include required containment isolation failures allowing a flow path to the environment. A release pathway outside primary containment exists when flow is not prevented by downstream isolations. In the case of a failure of both isolation valves to close but in which no downstream flowpath exists, emergency declaration under this threshold would not be required .

Similarly, if the emergency response requires the normal process flow of a system outside containment (e.g., EOP requirement to bypass MSIV low RPV water level interlocks and maintain the main condenser as a heat sink using main turbine bypass valves), the threshold is not met. The combination of these threshold conditions represent the loss of both the RCS and Containment (see PC Loss E Threshold #1) barriers and justifies declaration of a Site Area Emergency (i.e., Loss or Potential Loss of any two barriers). (ref. 1-4)

Even though RWCU and Feedwater systems do not contain steam, they are included in the list because an unisolable break could result in the high-pressure discharge of fluid that is flashed to steam from relatively large volume systems directly connected to the RCS. (ref. 1)

Large high-energy lines that rupture outside primary containment can discharge significant amounts of inventory and jeopardize the pressure-retaining capability of the RCS until they are isolated. If it is determined that the ruptured line cannot be promptly isolated from the control room, the RCS barrier Loss threshold is met.

CGS Basis Reference(s):

1. FSAR Section 5.4.5

2. FSAR Section 5.4.6

3. FSAR Section 5.4.8

4. FSAR Section 10.3

5. NEI 99-01 RCS Leak Rate RCS Loss 3.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency -Technical Bases Page: 79 of 190 Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

Emergency RPV Depressurization is required Basis:

Plant symptoms requiring Emergency RPV Depressurization per the EOPs are indicative of a loss of the RCS barrier. Emergency RPV Depressurization is specified in the EOP flowcharts when symbols containing the phrase "EMERG DEPRESS REQ'D" are reached (ref. 1-7). If Emergency RPV Depressurization is required, the plant operators are directed to open safety relief valves (SRVs) and keep them open as needed to maintain adequate core cooling with available injection sources (ref. 8, 9). Even though the RCS is being vented into the suppression pool, a loss of the RCS exists due to the diminished effectiveness of the RCS pressure barrier to a release of fission products beyond its boundary.

• Emergency RPV Depressurization in accordance with the EOPs is indicative of a loss of the RCS barrier. If Emergency RPV Depressurization is performed, the plant operators are directed to open safety relief valves (SRVs) and keep them open. Even though the RCS is being vented into the suppression pool, a Loss of the RCS barrier exists due to the diminished effectiveness of the RCS to retain fission products within its boundary.

CGS Basis Reference(s):

1. PPM 5.1.1 RPV Control

2. PPM 5.1.2 RPV Control - ATWS

3. PPM 5.1.4 RPV Flooding

4. PPM 5.1.6 RPV Flooding -ATWS

5. PPM 5.2.1 Primary Containment Control

6. PPM 5.3.1 Secondary Containment Control

7. PPM 5.4.1 Radioactivity Release Control

8. PPM 5.1.3 Emergency RPV Depressurization

9. PPM 5.1.5 Emergency RPV Depressurization - ATWS

10. NEI 99-01 RCS Leak Rate RCS Loss 3.B

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency-Technical Bases Page: 80 of 190 Barrier: Reactor Coolant System Category: B. RCS Leak Rate Degradation Threat: Potential Loss Threshold:

UNISOLABLE primary system leakage that results in exceeding EITHER:

RB area temperature alarm level (PPM 5.3.1 Table 23)

OR RB area radiation alarm level (PPM 5.3.1 Table 24)

Basis:

The presence of elevated general area temperatures or radiation levels in the secondary containment may be indicative of unisolable primary system leakage outside the primary containment. The PPM 5.3.1 Table 23 and Table 24 alarm levels define this RCS threshold because they are the maximum normal operating values and signify the onset of abnormal system operation. When parameters reach this level, equipment failure or misoperation may be occurring. Elevated parameters may also adversely affect the ability to gain access to or operate equipment within the affected area. The locations into which the primary system discharge is of concern correspond to the areas addressed in PPM 5.3.1 Tables 23 and 24 (ref. 1).

Area temperature alarms are provided by the leak detection and reactor building recirculation air (RRA) systems (ref. 2)

The ARM alarm setpoints listed in Table 24 vary due to plant operating mode and Health Physics radiation surveys. A program is established to maintain the current setpoint values in PPM 4.602.A5 for annunciator window 3-1; thus, reference is made to the annunciator response procedure in Table 24. '

(ref. 2)

In general, multiple indications should be used to determine if a primary system is discharging outside primary containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment.

Potential loss of RCS based on primary system leakage outside the primary containment is determined from EOP temperature or radiation Max Normal Operating values in areas such as main steam line tunnel, RCIC, etc., which indicate a direct path from the RCS to areas outside primary containment.

A Max Normal Operating value is the highest value of the identified parameter expected to occur during normal plant operating conditions with all directly associated support and control systems functioning properly. *

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 81 of 190 The indicators reaching the threshold barriers and confirmed to be caused by RCS leakage from a primary system warrant an Alert classification. A primary system is defined to be the pipes, valves, and other equipment which connect directly to .the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

An UNISOLABLE leak which is indicated by Max Normal Operating values escalates to a Site Area Emergency when combined with Containment Barrier Loss threshold 3.A (after a containment isolation) and a General Emergency when the Fuel Clad Barrier criteria is also exceeded.

CGS Basis Reference(s):

1. PPM 5.3.1 Secondary Containment Control

2. PPM 5.0.10 Flowchart Training Manual

3. NEI 99-01 RCS Leak Rate RCS Potential Loss 3.A

• Attachment 6.1, EAL Technical Bases

1*

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 82 of 190 Barrier: Reactor Coolant System Category: C. PC Conditions Degradation Threat: Loss Threshold:

PC pressure GT 1.68 psig due to RCS leakage Basis:

The drywell high pressure scram setpoint is an entry condition to the EOP flowcharts: PPM 5.1.1, RPV Control, and PPM 5.2.1, Primary Containment Control (ref. 1, 2, 3). Normal primary containment (PC) pressure control functions such as operation of drywell cooling and venting through SGT are specified in PPM 5.2.1 in advance of less desirable but more effective functions such as operation of drywell or wetwell sprays.

In the CGS design basis, primary containment pressures above the drywell high pressure scram setpoint are assumed to be the result of a high-energy release into the containment for which normal pressure control systems are inadequate or incapable of reversing the increasing pressure trend.

Pressures of this magnitude, however, can be caused by non-LOCA events such as a loss of drywell cooling or inability to control primary containment vent/purge (ref. 3).

The threshold phrase " ... due to RCS leakage" focuses the barrier failure on the RCS instead of the non-LOCA malfunctions that may adversely affect primary containment pressure. Primary containment pressure greater than 1.68 psig with corollary indications (e.g., elevated drywell temperature, indications of loss of RCS inventory) should therefore be considered a Loss of the RCS barrier. Loss of drywell cooling that results in pressure greater than 1.68 psig should not be considered an RCS barrier loss.

1.68 psig is the drywell high pressure setpoint which indicates a LOCA by automatically initiating the EGGS or equivalent makeup system.

There is no Potential Loss threshold associated with drywell pressure.

CGS Basis Reference(s):

1. Technical Specifications Table 3.3.5.1-1

2. PPM 5.1.1 RPV Control

3. PPM 5.2.1 Primary Containment Control

4. FSAR Section 6

5. NEI 99-01 Primary Containment Pressure RCS Loss 1.A

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 83 of 190 Barrier: Reactor Coolant System Category: D. PC Radiation/ RCS Activity Degradation Threat: Loss Threshold:

Containment Radiation Monitor CMS-RIS-27E or CMS-RIS-27F reading GT 70 R/hr Basis:

Four high range area radiation detectors (CMS-RE-27A, 8, E and F) are installed in the drywell. CMS-RE-27A and -278 are located in the bioshield wall at elevations 522' and 525', azimuth 60° and 297°,

respectively. CMS-RE-27E and -27F are located inside containment at elevation 515', azimuth 290° and 51.5°, respectively. The companion containment radiation monitors (CMS-RIS-27A, B, E and F) are located on RAD Boards 22 and 23 in the Main Control Room. (ref. 1)

The threshold value was calculated assurning the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with normal operating concentrations (i.e., within Technical Specifications) into the drywell atmosphere. Evaluation of detector location, geometry and

• anticipated response suggests CMS-RIS-27E or F will provide the desired response to a given radiation source in the drywell and are, therefore, identified as the p,referred monitors for evaluating this RCS Loss threshold. (ref. 2)

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that reactor coolant activity equals Technical Specification allowable limits. This value is lower than that specified for Fuel Clad Barrier Loss threshold D.1 since it indicates a loss of the RCS Barrier only.

There is no Potential Loss threshold associated with primary containment radiation.

CGS Basis Reference(s):

1. TM-2117 TSG - Core Thermal Engineer, Attachment 4.2

2. Calculation NE-02-94-57

3. NEI 99-01 Primary Containment Radiation RCS Loss 4.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 84 of 190 Barrier: Reactor Coolant System Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold:

Any condition in the opinion of the Emergency Director that indicates loss of the RCS barrier Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining if the RCS barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to 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 from portable instrumentation and consideration of offsite monitoring results .

Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is lost.

CGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Loss 6.A

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 85 of 190 Barrier: Reactor Coolant System Category: F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:

Any condition in the opinion of the Emergency Director that indicates potential loss of the RCS barrier Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining if the RCS barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to 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 from portable instrumentation and consideration of offsite monitoring results.

Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the RCS Barrier is potentially lost. The Emergency Director should also consider whether or not to declare the barrier potentially lost in the event that barrier status cannot be monitored.

CGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment RCS Potential Loss 6.A

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 86 of 190 Barrier: Containment Category: A. RPV Water Level Degradation Threat: Potential Loss Threshold:

\ SAG entry required Basis:

EOP flowcharts provide instructions to assure adequate core cooling by restoring and maintaining RPV water level above prescribed limits, operate sufficient RPV injection sources to assure adequate core cooling, and assess the possibility of core damage when RPV level cannot be determined. The Fuel Clad Loss threshold conditions are the EOP flowchart conditions that signal a loss of adequate core cooling and a requirement to exit all EOPs and enter the SAGs (ref. 1-6).

This threshold is also a Loss of the RCS barrier (RCS Loss A) and a Loss of the Fuel Clad barrier (FC Loss A), and therefore represents a Loss of two barriers and a Potential Loss of a third, which requires a General Emergency classification.

The Potential Loss threshold is identical to the Fuel Clad Loss RPV Water Level threshold. The Potential Loss requirement for SAG entry indicates adequate core cooling cannot be restored and maintained and that core damage is possible. BWR EPGs/SAGs specify the conditions that require SAG entry. When SAG entry is required, the EPGs are exited and SAGs are entered. Entry into SAGs is a logical escalation in response to the inability to restore and maintain adequate core cooling.

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

CGS Basis Reference(s):

1. PPM 5.1.1 RPV Control

2. PPM 5.1.2 RPV Control - A TWS

3. Calculation NE-02-03-06 Attachment 10 RPV Variables

4. PPM 5.0.10 Flowchart Training Manual

5. PPM 5.1.4 RPV Flooding

6. PPM 5.1.6 RPV Flooding - A TWS

7. NEI 99-01 RPV Water Level PC Potential Loss 2.A

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 87 of 190 Barrier: Containment Category: B. RCS Leak Rate Degradation Threat: Loss Threshold:

UNISOLABLE primary system leakage that results in exceeding EITHER:

RB area maximum safe operating temperature (PPM 5.3.1 Table 23)

OR RB area maximum safe operating radiation (PPM 5.3.1 Table 24)

Basis:

The presence of elevated general area temperatures or radiation levels in the Reactor Building (RB) may be indicative of unisolable primary system leakage outside the primary containment. The maximum safe operating values define this Containment barrier threshold because they are indicative of problems in the secondary containment that are spreading and pose a threat to achieving a safe plant shutdown. This threshold addresses problematic discharges outside primary containment that may not originate from a high-energy line break. The locations into which the primary system discharge is of concern correspond to the areas addressed in PPM 5.3.1 Tables 23 and 24 (ref. 1).

RB maximum safe operating temperatures are conservatively defined by the qualification temperature of safety related equipment in the area. The equipment qualification program has proven that safety related equipment will perform satisfactorily to at least this temperature. In an area with multiple components and different qualification temperatures, the maximum safe operating temperature assigned to that area is generally the lowest of the individual temperatures. (ref. 2)

The maximum safe operating radiation value is defined to be 10,000 mR/hr in areas other than the refueling floor. This is the maximum indication on all but the high level instruments. This value is high enough to be indicative of substantial and immediate problems yet low enough to allow time for shutdown or isolation of a leak without exceeding the total integrated dose allowable for even the most sensitive safety related equipment. No area radiation levels are defined for the refueling floor because no primary systems are routed there. (ref. 2)

In general, multiple indications should be used to determine if a primary system is discharging outside primary containment. For example, a high area radiation condition does not necessarily indicate that a primary system is discharging into the secondary containment since this may be caused by *radiation shine from nearby steam lines or the movement of radioactive materials. Conversely, a high area radiation condition in conjunction with other indications (e.g. room flooding, high area temperatures, reports of steam in the secondary containment, an unexpected rise in feedwater flowrate, or unexpected main turbine control valve closure) may indicate that a primary system is discharging into the secondary containment.

The Max Safe Operating Temperature and the Max Safe Operating Radiation Level are each the highest value of these parameters at which neither: (1) equipment necessary for the safe shutdown of the plant will fail, nor (2) personnel access necessary for the safe shutdown of the plant will be precluded. EOPs utilize these temperatures and radiation levels to establish conditions under which RPV depressurization is required.

Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 88 of 190 The temperatures and radiation levels should be confirmed to be caused by RCS leakage from a primary system. A primary system is defined to be the pipes, valves, and other equipment which connect directly to the RPV such that a reduction in RPV pressure will effect a decrease in the steam or water being discharged through an unisolated break in the system.

In combination with the RCS Potential Loss RCS Leak Rate threshold this threshold would result in a Site Area Emergency.

There is no Potential Loss threshold associated with primary containment isolation failure.

CGS Basis Reference(s):

1. PPM 5.3.1 Secondary Containment Control

2. PPM 5.0.10 Flowchart Training Manual

3. NEI 99-01 RCS Leak Rate PC Loss 3.C

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 89 of 190 Barrier: Containment Category: C. PC Conditions Degradation Threat: Loss Threshold:

UNPLANNED rapid drop in PC pressure following PC pressure rise Basis:

Rapid UNPLANNED loss of primary containment pressure (i.e., not attributable to containment spray or condensation effects) following an initial pressure increase indicates a loss of primary containment integrity.

This threshold relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.

CGS Basis Reference(s):

1. NEI 99-01 Primary Containment Conditions PC Loss 1.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A 1*use Category: REFERENCE Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 90 of 190 Barrier: Containment Category: C. PC Conditions Degradation Threat: Loss Threshold:

PC pressure response not consistent with LOCA conditions Basis:

This indicator is considered to be a loss of both the RCS and PC barriers.

Normal LOCA conditions are drywell pressure rising with wetwell pressure following. Primary containment or drywell pressure responses not consistent with LOCA conditions indicate a loss of the primary containment barrier. This may be noticed as a decrease in drywell pressure when no operator action (e.g., starting drywell cooling fans) has been taken. It would also include a failure of the drywell pressure to increase as expected during a LOCA. Also, a loss of suppression function in conjunction with a LOCA would indicate a loss of the primary containment barrier. Exceeding Pressure Suppression Pressure (PSP) is an indication of loss of pressure suppression function .

• Primary containment pressure should increase as a result of mass and *energy release into the primary containment from a LOCA. Thus, primary containment pressure not increasing under these conditions indicates a loss of primary containment integrity.

This threshold relies on operator recognition of an unexpected response for the condition and therefore a specific value is not assigned. The unexpected (UNPLANNED) response is important because it is the indicator for a containment bypass condition.

CGS Basis Reference(s):

1. FSAR Section 6.2.1.1.3.3

2. FSAR Figure 6.2-3

3. FSAR Table 6.2-5

4. FSAR Table 6.2-1

5. NEI 99-01 Primary Containment Conditions PC Loss 1.B

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 91 of 190 Barrier: Containment Category: C. PC Conditions Degradation Threat: Potential Loss Threshold:

I PC pressure GT 45 psig Basis:

If this threshold is exceeded, a challenge to the primary containment structure has occurred because assumptions used in the accident analysis are no longer valid and an unanalyzed condition exists (ref. 1, 2). This constitutes a Potential Loss of the Containment barrier even if a containment breach has not occurred.

The threshold pressure is the primary containment internal design pressure. Structural acceptance testing demonstrates the capability of the primary containment to resist pressures greater than the internal design pressure. A pressure of this magnitude is greater than those expected to result from any design basis accident and, thus, represent a Potential Loss of the Containment barrier.

CGS Basis Reference(s):

1. FSAR Table 6.2-1

2. FSAR Section 6.2

3. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A J Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 92 of 190 Barrier: Containment Category: C. PC Conditions Degradation Threat: Potential Loss Threshot'd:

Explosive mixture exists inside PC (H 2 GE 6% and 02 GE 5%)

Basis:

Explosive (deflagration) mixtures in the primary containment are assumed to be elevated concentrations of hydrogen and oxygen. BWR industry evaluation of hydrogen generation for development of EOPs/SAGs indicates that any hydrogen concentration above minimum detectable is not to be expected within the short term. Post-LOCA hydrogen generation primarily caused by radiolysis is a slowly evolving, long-term condition. Hydrogen concentrations that rapidly develop are most likely caused by metal-water reaction. A metal-water reaction is indicative of an accident more severe than accidents considered in *the plant design basis and would be indicative, therefore, of a potential threat to primary containment integrity. Hydrogen concentration of approximately 6% is considered the global deflagration concentration limit (ref. 1) .

• Except for brief periods during plant startup and shutdown, oxygen concentration in the primary containment is maintained at insignificant levels by nitrogen inerting. The specified values for this Potential Loss threshold are the minimum global deflagration concentration limits (6% hydrogen and 5% oxygen, ref. 2) and readily recognizable because 6% hydrogen is well above the EOP flowchart entry condition (ref. 3). The,minimum global deflagration hydrogen/oxygen concentrations (6% / 5%,

respectively) require intentional primary containment venting, which is defined to be a Loss of Containment (PC Integrity or Bypass).

If hydrogen concentration reaches or exceeds the lower flammability limit, as defined in plant EOPs, in an oxygen rich environment, a potentially explosive mixture exists. If the combustible mixture ignites inside the primary containment, loss of the Containment barrier could occur.

CGS Basis Reference(s):

1. BWROG EPG/SAG Revision 2, Sections PC/G

2. PPM 5.7.1 RPV and Primary Containment Flooding SAG, Table 19

3. PPM 5.2.1 Primary Containment Control

4. FSAR Section 7.5.1.5.4

5. PPM 5.0.1 O Flowchart Training Manual

6. PPM 4.814.J1 814.J1 Annunciator Panel Alarms, 2-2

7. PPM 4.814.J2 814.J2 Annunciator Panel Alarms, 2-2

8. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.8

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency- Technical Bases Page: 93 of 190 Barrier: Containment Category: C. PC Conditions Degradation Threat: Potential Loss Threshold:

WW temperature and RPV pressure cannot be maintained below the HCTL Basis:

The HCTL is given in EOP flowchart Figure C (ref. 1). This is the only instance in which the threshold could be met.

Heat Capacity Temperature Limit (HCTL) is the highest Wetwell temperature from which emergency RPV depressurization will not exceed:

• Capability of the Wetwell, and equipment within the Wetwell which may be required to operate, when the RPV is pressurized

• Pressure Limit (PCPL), while the rate of energy transfer from the RPV to the Containment is GT the capacity of the Containment vent

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

CGS Basis Reference(s):

1. PPM 5.2.1 Primary Containment Control

2. NEI 99-01 Primary Containment Conditions PC Potential Loss 1.C

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency -Technical Bases Page: 94 of 190 Barrier: Containment -

Category: D. PC Radiation / RCS Activity Degradation Threat: Potential Loss Threshold:

Containment Radiation Monitor CMS-RIS-27E or CMS-RIS-27F reading GT 14,000 R/hr Basis:

Four high range area radiation detectors (CMS-RE-27A, B, E and F) are installed in the drywell.

CMS-RE-27A and -27B are located in the bioshield wall at elevations 522' and 525', azimuth 60° and 297°, respectively. CMS-RE-27E and -27F are located inside containment at elevation 515', azimuth 290° and 51.5°, respectively. The companion containment radiation monitors (CMS-RIS-27A, B, E and F) are located on RAD Boards 22 and 23 in the Main Control Room. (ref. 1)

The threshold value was calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with 20% fuel clad damage into the drywell atmosphere. Evaluation of detector location, geometry and anticipated response suggests CMS-RIS-27E or F will provide the desired response to a given radiation source in the drywell and are, therefore, identified as the preferred monitors for evaluating this Containment barrier Potential Loss threshold.

(ref. 2)

The radiation monitor reading corresponds to an instantaneous release of all reactor coolant mass into the primary containment, assuming that 20% of the fuel cladding has failed. This level of fuel clad failure is well above that used to determine the analogous Fuel Clad Barrier Loss and RCS Barrier Loss thresholds.

NUREG-1228, Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents, indicates the fuel clad failure must be greater than approximately 20% in order for there to be a major release of radioactivity requiring offsite protective actions. For this condition to exist there must already have been a loss of the RCS Barrier and the Fuel Clad Barrier. It is therefore prudent to treat this condition as a potential loss of containment which would then escalate the emergency classification level to a General Emergency.

CGS Basis Reference(s):

1. TM-2117 TSG - Core Thermal Engineer, Attachment 4.2

2. Calculation NE-02-94-57

3. NEI 99-01 Primary Containment Radiation Fuel Clad Potential Loss 1.D

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 95 of 190 Barrier: Containment Category: E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

UNISOLABLE direct downstream pathway to the environment exists after PC isolation signal Basis:

This threshold addresses failure of open isolation devices which should close upon receipt of a manual or automatic containment isolation signal resulting in a significant radiological release pathway directly to the environment. The concern is the unisolable open pathway to the environment. A failure of the ability to isolate any one line indicates a breach of containment integrity.

• Leakage into a closed system is to be considered only if the closed system is breached and thereby creates a significant pathway to the environment. Examples include unisolable main steam line or RCIC steam line breaks, unisolable RWCU system breaks, and unisolable PC vent paths.

PPM 5.2.1, Primary Containment Control, may specify primary containment venting and intentional

• bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). Under these conditions with a valid containment isolation signal, the Containment barrier should be considered lost.

The use of the modifier "direct" in defining the release path discriminates against release paths through interfacing liquid systems or minor release pathways, such as instrument lines, not protected by the Primary Containment Isolation System (PCIS).

The existence of a filter is not considered in the threshold assessment. Filters do not remove fission product noble gases. In addition, a filter could become ineffective due to iodine and/or particulate loading beyond design limits (i.e., retention ability has been exceeded) or water saturation from steam/high humidity in the release stream.

Following the leakage of RCS mass into primary containment and a rise in primary containment pressure, there may be minor radiological releases associated with allowable primary containment leakage through various penetrations or system components. Minor releases may also occur if a primary containment isolation valve(s) fails to close but the primary containment atmosphere escapes to an enclosed system. These releases do not constitute a loss or potential loss of primary containment but should be evaluated using the Recognition Category R ICs.

CGS Basis Reference(s):

1. PPM 5.2.1 Primary Containment Control

2. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.A

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 96 of 190 Barrier: Containment Category: E. PC Integrity or Bypass Degradation Threat: Loss Threshold:

[ Intentional PC venting per EOPs Basis:

EOP flowcharts (PPM 5.2.1, Primary Containment Control) may specify primary containment venting and intentional bypassing of the containment isolation valve logic, even if offsite radioactivity release rate limits are exceeded (ref. 1). The threshold is met when the operator begins venting the primary containment in accordance with EOP Support Procedures (PPM 5.5.14 or PPM 5.5.15) or ABN-CONT-VENT, not when actions are taken to bypass interlocks prior to opening the vent valves (ref. 2, 3, 4).

Purge and vent actions specified in PPM 5.2.1 to control primary containment pressure below the drywell high pressure scram setpoint or to lower hydrogen concentration does not meet this threshold because such action is only permitted if offsite radioactivity release rates will remain below the ODCM RFO limits (ref. 1) .

EOPs may direct primary containment isolation valve logic(s) to be intentionally bypassed, even if offsite radioactivity release rate limits will be exceeded. Under these conditions with a valid primary containment isolation signal, the containment should also be considered lost if primary containment venting is actually performed.

Intentional venting of primary containment for primary containment pressure control to the secondary containment and/or the environment is a Loss of the Containment. Venting for primary containment pressure control when not in an accident situation (e.g., to control pressure below the drywell high pressure scram setpoint) does not meet the threshold condition.

CGS Basis Reference(s):

1. PPM 5.2.1 Primary Containment Control

2. PPM 5.5.14 Emergency Wetwell Venting

3. PPM 5.5.15 Emergency Drywell Venting

4. ABN-CONT-VENT

5. NEI 99-01 Primary Containment Isolation Failure PC Loss 3.8

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 97 of 190 Barrier: Containment Category: F. Emergency Director Judgment Degradation Threat: Loss Threshold:

Any condition in the opinion of the Emergency Director that indicates loss of the Containment barrier Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining if the Containment barrier is lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to 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 from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and ATWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.

CGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency- Technical Bases Page: 98 of 190 Barrier: Containment Category: F. Emergency Director Judgment Degradation Threat: Potential Loss Threshold:

Any condition in the opinion of the Emergency Director that indicates potential loss of the Containment barrier Basis:

The Emergency Director judgment threshold addresses any other factors relevant to determining if the Containment barrier is potentially lost. Such a determination should include imminent barrier degradation, barrier monitoring capability and dominant accident sequences.

• Imminent barrier degradation exists if the degradation will likely occur within two hours based on a projection of current safety system performance. The term "imminent" refers to 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 from portable instrumentation and consideration of offsite monitoring results.

• Dominant accident sequences lead to degradation of all fission product barriers and likely entry to the EOPs. The Emergency Director should be mindful of the Loss of AC power (Station Blackout) and A TWS EALs to assure timely emergency classification declarations.

This threshold addresses any other factors that may be used by the Emergency Director in determining whether the Containment Barrier is lost.

CGS Basis Reference(s):

1. NEI 99-01 Emergency Director Judgment PC Loss 6.A

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency- Technical Bases Page: 99 of 190 Category H - Hazards and Other Conditions Affecting Plant Safety EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Hazards are non-plant, system-related events that can directly or indirectly affect plant operation, reactor plant safety or personnel safety.

1. Security Unauthorized entry attempts into the Protected Area, bomb threats, sabotage attempts, and actual security compromises threatening loss of physical control of the plant.

2. Seismic Event Natural events such as earthquakes have potential to cause plant structure or equipment damage of sufficient magnitude to threaten personnel or plant safety.

3. Natural or Technology Hazard Other natural and non-naturally occurring events that can cause damage to plant facilities include tornados, FLOODING, hazardous material releases and events restricting site access warranting classification .

• 4. Fire Fires can pose significant hazards to personnel and reactor safety. Appropriate for classification are fires within the site Protected Area or which may affect operability of equipment needed for safe shutdown.

5. Hazardous Gas Toxic, corrosive, asphyxiant or flammable gas leaks can affect normal plant operations or preclude access to plant areas required to safely shutdown the plant.

6. Control Room Evacuation If the Control Room must be evacuated, additional support for monitoring and controlling plant functions is necessary through the emergency response facilities.

7. Emergency Director Judgment The EALs defined in other categories specify the predetermined symptoms or events that are indicative of emergency or potential emergency conditions and thus warrant classification. While these EALs have been developed to address the full spectrum of possible emergency conditions which may warrant classification and subsequent implementation of the Emergency Plan, a provision for classification of emergencies based on operator/management experience and judgment is still necessary. The EALs of this category provide the Emergency Director the latitude to classify emergency conditions consistent with the established classification criteria based upon Emergency Director judgment.

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 100 of 190 Category: H -Hazards Subcategory: 1 -Security Initiating Condition: HOSTILE ACTION within the Protected Area EAL:

HS1.1 Site Area Emergency A HOSTILE ACTION is occurring or has occurred within the PROTECTED AREA as reported by the Security Sergeant or Security Lieutenant Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

The Security Shift Supervision is defined as either the Security Lieutenant or the Security Sergeant (ref. 1).

Note that the ISFSI Protected Area is an area separate from the Protected Area surrounding the power block.

The Safeguards Contingency Plan (Appendix C of CGS Physical Security Plan) defines the events that meet the criteria of a SECURITY CONDITION or HOSTILE ACTION (ref. 1) .

This IC addresses the occurrence of a HOSTILE ACTION within the PROTECTED AREA. This event will require rapid response and assistance due to the possibility for damage to plant equipment.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Site Area Emergency declaration will mobilize Offsite Response Organization (ORO) resources and have them available to develop and implement public protective actions in the unlikely event that the attack is successful in impairing multiple safety functions.

This IC does not apply to a HOSTILE ACTION directed at an ISFSI PROTECTED AREA located outside the plant PROTECTED AREA; such an attack should be assessed using IC HA 1. It also does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73.71 or 10 CFR § 50.72.

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the CGS Physical Security Plan (ref. 1) .

CGS Basis Reference(s):

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 101 of 190

1. CGS Physical Security Plan

2. NEI 99-01 HS1

• Attachment 6.1, EAL Technical Bases

I Use Category:

Number: 13.1.1A REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 102 of 190 Category: H-Hazards Subcategory: 1 -Security Initiating Condition: HOSTILE ACTION within the OWNER CONTROLLED AREA or airborne attack threat within 30 minutes EAL:

HA1.1 Alert (1) A HOSTILE ACTION is occurring or has occurred within the OWNER CONTROLLED AREA as reported by the Security Sergeant or Security Lieutenant OR (2) A validated notification from NRC of an aircraft attack threat within 30 min. of the site Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

The Security Shift Supervision is defined as either the Security Lieutenant or the Security Sergeant (ref. 1).

• Note that the ISFSI Protected Area is an area separate from the Protected Area surrounding the power block.

The Safeguards Contingency Plan (Appendix C of CGS Physical Security Plan) defines the events that meet the criteria of a SECURITY CONDITION or HOSTILE ACTION (ref. 1).

This IC addresses the occurrence of a HOSTILE ACTION within the OWNER CONTROLLED AREA or notification of an aircraft attack threat. This event will require rapid response and assistance due to the possibility of the attack progressing to the PROTECTED AREA, or the need to prepare the plant and staff for a potential aircraft impact.

Timely and accurate communications between the Security Shift Supervision and the Control Room is essential for proper classification of a security-related event.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

As time and conditions allow, these events require a heightened state of readiness by the plant staff and implementation of onsite protective measures (e.g., evacuation, dispersal or sheltering). The Alert declaration will also heighten the awareness of Offsite Response Organizations (OROs), allowing them to be better prepared should it be necessary to consider further actions.

This IC does not apply to incidents that are accidental events, acts of civil disobedience, or otherwise are not a HOSTILE ACTION perpetrated by a HOSTILE FORCE. Examples include the crash of a small aircraft, shots from hunters, physical disputes between employees, etc. Reporting of these types of events is adequately addressed by other EALs, or the requirements of 10 CFR § 73. 71 or 10 CFR § 50.72 .

• Threshold #1 is applicable for any HOSTILE ACTION occurring, or that has occurred, in the OWNER CONTROLLED AREA. This includes any action directed against the ISFSI which is located outside the plant PROTECTED AREA.

Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor' Rev: N/A

Title:

Classifying The Emergency - Tec~nical Bases Page: 103 of 190 Threshold #2 addresses the threat from the impact of an aircraft on the plant, and the anticipated arrival time is within 30 minutes. The 'intent of this EAL is to ensure that threat-related notifications are made in a timely manner so that plant personnel and OROs are in a heightened state of readiness. This EAL is met when the threat-related information has been validated in accordance with ABN-AIRBORNE-ATTACK (ref. 2).

The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may be provided by NORAD through the NRC.

In some cases, it may not be readily apparent if an aircraft impact withi.n the OWNER CONTROLLED AREA was intentional (i.e., a HOSTILE ACTION). It is expected, although not certain, that notification by an appropriate Federal agency to the site would clarify this point. In this case, the appropriate federal agency is intended to be NORAD, FBI, FAA or NRC. The emergency declaration, including one based on other ICs/EALs, should not be unduly delayed while awaiting notification by a Federal agency.

Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the CGS Physical Security Plan (ref. 1).

CGS Basis Reference(s):

• 1.

2.

3.

CGS Physical Security Plan ABN-AI RBORNE-ATTACK NEI 99-01 HA 1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 104 of 190 Category: H- Hazards Subcategory: 1 -Security Initiating Condition: Confirmed SECURITY CONDITION or threat EAL:

HU1 .1 Unusual Event (1) A SECURITY CONDITION that does not involve a HOSTILE ACTION as reported by the Security Sergeant or Security Lieutenant OR (2) Notification of a credible security threat directed at the site OR (3) A validated notification from the NRC providing information of an aircraft threat Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

The Security Shift Supervision is defined as either the Security Lieutenant or the Security Sergeant (ref. 1).

This EAL is based on the CGS Physical Security Plan (ref. 1).

The Safeguards Contingency Plan (Appendix C of CGS Physical Security Plan) defines the events that meet the criteria of a SECURITY CONDITION or HOSTILE ACTION (ref. 1).

This IC addresses events that pose a threat to plant personnel or SAFETY SYSTEM equipment, and thus represent a potential degradation in the level of plant safety. Security events which do not meet one of these EALs are adequately addressed by the requirements of 10 CFR § 73. 71 or 10 CFR § 50.72. Security events assessed as HOSTILE ACTIONS are classifiable under ICs HA1 and HS1.

Timely and accurate communications between Security Shift Supervision and the Control Room is essential for proper classification of a security-related event. Classification of these events will initiate appropriate threat-related notifications to plant personnel and Offsite Response Organizations.

Security plans and terminology are based on the guidance provided by NEI 03-12, Template for the Security Plan, Training and Qualification Plan, Safeguards Contingency Plan [and Independent Spent Fuel Storage Installation Security Program].

Threshold #1 references the Security Shift Supervision because these are the individuals trained to confirm that a security event is occurr:ing or has occurred. Training on security event confirmation and classification is controlled due to the nature of Safeguards and 10 CFR § 2.39 information.

Threshold #2 addresses the receipt of a credible security threat. The credibility of the threat is assessed in accordance with the CGS Physical Security Plan (ref. 1).

Threshold #3 addresses the threat from the impact of an aircraft on the plant. The NRC Headquarters Operations Officer (HOO) will communicate to the licensee if the threat involves an aircraft. The status and size of the plane may also be provided by NORAD through the NRC. Validation of the threat is performed in accordance with ABN-AIRBORNE-ATTACK (ref. 2).

Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency -Technical Bases Page: 105 of 190 Emergency plans and implementing procedures are public documents; therefore, EALs should not incorporate Security-sensitive information. This includes information that may be advantageous to a potential adversary, such as the particulars concerning a specific threat or threat location. Security-sensitive information should be contained in non-public documents such as the CGS Physical Security Plan (ref. 1).

Escalation of the emergency classification level would be via IC HA 1.

CGS Basis Reference{s):

1. CGS Physical Security Plan

2. ABN-AIRBORNE-ATTACK

3. NEI 99-01 HU1

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 106 of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 2 - Seismic Event Initiating Condition: Seismic event GT OBE levels EAL:

HU2.1 Unusual Event Seismic event GT Operating Basis Earthquake (OBE) as indicated by H13.P851.S1 .5-1 (OPERATING BASIS EARTHQUAKE EXCEEDED) activated Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

CGS seismic instrumentation consists of a Kinemetrics SMA-3 Strong Motion Accelerograph and associated sensors that are equipped with seismic triggers set to initiate recording at an acceleration equal to or exceeding 0.01 g (ref. 1, 2). This also annunciates the seismic activity alarm H13.P851.S1.2-5 Minimum Seismic Earthquake Exceeded (ref. 2, 3, 4).

A seismic switch unit that is similar to the seismic trigger unit is also provided. The trip point of the seismic switch unit is set at the maximum acceleration corresponding to the OBE, and it provides immediate Control Room annunciation that the OBE has been exceeded requiring declaration of an Unusual Event (ref. 1, 3, 4)

This IC addresses a seismic event that results in accelerations at the plant site greater than those specified for an Operating Basis Earthquake (OBE). An earthquake greater than an OBE but less than a Safe Shutdown Earthquake (SSE) should have no significant impact on safety-related systems, structures and components; however, some time may be required for the plant staff to ascertain the actual post-event condition of the plant (e.g., performs walk-downs and post-event inspections). Given the time necessary to perform walk-downs and inspections, and fully understand any impacts, this event represents a potential degradation of the level of safety of the plant.

Event verification with external sources should not be necessary during or following an OBE.

Earthquakes of this magnitude should be readily felt by on-site personnel and recognized as a seismic event (e.g., lateral accelerations in excess of 0.08g). The Shift Manager or Emergency may seek external verification if deemed appropriate (e.g., a call to the USGS, check internet news sources, etc.);

however, the verification action must not preclude a timely emergency declaration.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or MAB.

CGS Basis Reference(s):

1. CGS FSAR Section 3.7.4 Seismic Instrumentation

2. ISP-SEIS-M201 Seismic Systems Channel Check

3. PPM 4.851.S1 .2-5 Minimum Seismic Earthquake Exceeded

4. ABN-EARTHQUAKE Earthquake

5. NEI 99-01 HU2

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 107 of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.1 Unusual Event (1) A tornado strike within the PROTECTED AREA OR (2) Volcanic ash fallout requiring plant shutdown Mode Applicability:

1 1 1 2 1 3 I 4 I s I def I Basis:

If damage is confirmed visually or by other in-plant indications, the event may be escalated to an Alert under EAL CA6.1 or MA8.1.

Threshold #1 A tornado striking (touching down) within the PROTECTED AREA warrants declaration of an Unusual Event regardless of the measured wind speed at the meteorological tower. A tornado is defined as a violently rotating column of air in contact with the ground and extending from the base of a thunderstorm: A dust devil is not a tornado.

Threshold #2 The potential for volcanic eruption exists in the Pacific Northwest. Heavy ash fall, such as that experienced at certain locations following the erl;Jption of Mt. St. Helens in 1980, could affect operation of plant equipment if precautionary measures are not taken. The design basis ash fall is projected for a 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> duration. Plant shutdown may be warranted, based on several individual criteria specified in ABN-ASH (ref. 1). This threshold is met when ABN-ASH requires plant shutdown.

This IC *addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

Threshold #1 addresses a tornado striking (touching down) within the PROTECTED AREA.

Threshold #2 addresses a volcanic ash fallout event.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, M orC.

CGS Basis Reference(s):

1. ABN-ASH Ash Fall

2. NEI 99-01 HU3

• Attachment 6.1, EAL Technical Bases J

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 108 of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3 - Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.2 Unusual Event Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY SYSTEM component needed for the current operating mode Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

An uncontrolled flooding event may pose a direct threat to safety-related equipment. As such, the potential exists for substantial degradation of the level of safety of the plant. One indication of FLOODING is indicated by ECCS room level alarms on P601 (ref. 1, 2).

• This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

This EAL addresses FLOOD ING of a building room or area that results in operators isolating power to a SAFETY SYSTEM component due to water level or other wetting concerns. Classification is also required if the water level or related wetting causes an automatic isolation of a SAFETY SYSTEM component from its power source (e.g., a breaker or relay trip). To warrant classification, operability of the affected component must be required by Technical Specifications for the current operating mode.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, More.

CGS Basis Reference(s):

1. Calculation ME 02-02-02 Reactor Building Flooding

2. Calculation ME 02-02-46, RB/RW/TB/DG Corridor Flooding

3. NEI 99-01 HU3

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A j Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 109 of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 3- Natural or Technology Hazard Initiating Condition: Hazardous event EAL:

HU3.3 Unusual Event (1) Movement of personnel within the PROTECTED AREA is IMPEDED due to an offsite event involving hazardous materials (e.g., an offsite chemical spill, 618-11 event or toxic gas release)

OR (2) A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles (Note 7)

Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents Mode Applicability:

I 1 I 2 I3 J 4 I s I def J Basis:

As used here, the term "offsite" is meant to be areas external to the PROTECTED AREA.

Threshold #1 includes an event at the 618-11 burial ground which would IMPEDE movement of personnel within the PROTECTED AREA.

Threshold #2 includes a range fire causing Hanford officials to limit vehicle access to the site. The origin of the hazardous event could be from on or off-site.

This IC addresses hazardous events that are considered to represent a potential degradation of the level of safety of the plant.

Threshold #1 addresses a hazardous materials event originating at an offsite location and of sufficient magnitude to impede the movement of personnel within the PROTECTED AREA.

Threshold #2 addresses a hazardous event that causes an on-site impediment to vehicle movement and significant enough to prohibit the plant staff from accessing the site using personal vehicles.

Examples of such an event include site flooding caused by a hurricane, heavy rains, up-river water releases, dam failure, etc., or an on-site train derailment blocking the access road.

This EAL is not intended apply to routine impediments such as fog, snow, ice, or vehicle breakdowns or accidents, but rather to more significant conditions such as the Hurricane Andrew strike on Turkey Point in 1992, the flooding around the Cooper Station during the Midwest floods of 1993, or the flooding around Ft. Calhoun Station in 2011.

Escalation of the emergency classification level would be based on ICs in Recognition Categories R, F, More.

CGS Basis Reference(s):

• 1. NEI 99-01 HU3 Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 110 of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4- Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.1 Unusual Event A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):

• Report from the field (i.e., visual observation)

• Receipt of multiple (more than 1) fire alarms or indications

• Field verification of a single fire alarm AND The FIRE is located within any Table 5 area Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 5 Safe Shutdown Areas

• Vital portions of the Rad Waste/Control Building:

467' elevation vital island 487' elevation cable spreading room Main Control Room and vertical cable chase 525' elevation HVAC area

• Reactor Building

• Vital portions of the Turbine Building DEH pressure switches RPS switches on turbine throttle valves Main steam line radiation monitors Turbine Building ventilation radiation monitors Main steam line piping up to MS-V-146 and the first stop valves

• Standby Service Water Pump Houses

• Diesel Generator Building Mode Applicability:

1 2 3 4 s I def 1 Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 111 of 190 Basis:

A fire alarm can be confirmed by multiple/redundant indications such as additional alarms on FCP-1 or FCP-2, fire pumps starting, fire suppression system discharge, fire water header pressure fluctuations or by notification by plant personnel (ref. 1).

The Table 5 Safe Shutdown Areas include those structures/areas that contain any Class 1, 2 or 3 SSC.

Table 5 includes those structures containing functions and systems required to achieve and maintain cold shutdown (including all auxiliary equipment such as AC/DC power, cooling water and instrumentation) (ref. 2).

• The concept of this EAL is that a fire exists in a Table 5 area that is not extinguished within 15 minutes.

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.

For EAL HU4.1 the intent of the 15-minute duration is to size the Fl RE and to discriminate against small FIRES that are readily extinguished (e.g., smoldering waste paper basket). In addition to alarms, other indications of a FIRE could be a drop in fire main pressure, automatic activation of a suppression system, etc.

Upon receipt, operators will take prompt actions to confirm the validity of an initial fire alarm, indication,*

or report. For EAL assessment purposes, the emergency declaration clock starts at the time that the initial alarms, indications, or report was received, and not the time that a subsequent verification action was performed. Similarly, the fire duration clock also starts at the time of receipt of the initial alarms, indications or report.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or MAB.

CGS Basis Reference(s):

1. ABN-FIRE

2. FSAR Table 3.2-1 Equipment Classification

3. NEI 99-01 HU4

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 112 of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4- Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.2 Unusual Event Receipt of a single fire alarm (i.e., no other indications of a FIRE)

AND The fire alarm is indicating a FIRE within any Table 5 area AND The existence of a FIRE is not verified within 30 min. of alarm receipt (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 5 Safe Shutdown Areas

• Vital portions of the Rad Waste/Control Building:

• 467' elevation vital island 487' elevation cable spreading room Main Control Room and vertical cable chase 525' elevation HVAC area

• Reactor Building

• Vital portions of the Turbine Building DEH pressure switches RPS switches on turbine throttle valves Main steam line radiation monitors Turbine Building ventilation radiation monitors Main steam line *piping up to MS-V-146 and the first* stop valves

• Standby Service Water Pump Houses

• Diesel Generator Building Mode Applicability:

1 2 3 4 s I def 1

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 113 of 190 Basis:

The 30 minute requirement begins upon receipt of a single valid 1 fire detection system alarm. The alarm is to be validated using available Control Room indications or alarms to prove that it is not spurious, or by reports from the field. Actual field reports must be made within the 30 minute time limit or a classification must be made. If a fire is verified to be occurring by field report, classification shall be made based on EAL HU4.1.

A single point fire alarm, with no other indications of a fire, may be more indicative of an instrumentation issue rather than a fire in the plant.

The concept of this EAL is that there is 30 minutes to determine if a fire exists when only one fire alarm

. is received.

The Table 5 Safe Shutdown Areas include those structures/areas that contain any Class 1, 2 or 3 SSC.

Table 5 includes those structures containing functions and systems required to achieve and maintain cold shutdown (including all auxiliary equipment such as AC/DC power, cooling water and instrumentation) (ref. 1).

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.

This EAL addresses receipt of a single fire alarm, and the existence of a FIRE is not verified (i.e.,

proved or disproved) within 30-minutes of the alarm. Upon receipt, operators will take prompt actions to confirm the validity of a single fire alarm. For EAL assessment purposes, the 30-minute clock starts at the time that the initial alarm was received, and not the time that a subsequent verification action was performed.

A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For !his reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.

• If an actual FIRE is verified by a report from the field, then HU4.1 is immediately applicable, and the emergency must be declared if the FIRE is not extinguished within 15-minutes of the report. If the alarm is verified to be due to an equipment failure or a spurious activation, and this verification occurs within 30-minutes of the receipt of the alarm, then this EAL is not applicable and no emergency declaration is warranted.

Basis-Related Requirements from Appendix R Appendix R to 10 CFR 50, states in part:

Criterion 3 of Appendix A to this part specifies that "Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions."

When considering the effects of fire, those systems associated with achieving and maintaining safe shutdown conditions assume major importance to safety because damage to them can lead to core damage resulting from loss of coolant through boil-off.

Because fire may affect safe shutdown systems and because the loss of function of systems used

• to mitigate the consequences of design basis accidents under post-fire conditions does not per se impact public safety, the need to limit fire damage to systems required to achieve and maintain safe Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 114 of 190 shutdown conditions is greater than the need to limit fire damage to those systems required to mitigate the consequences of design basis accidents.

In addition, Appendix R to 10 CFR 50, requires, among other considerations, the use of 1-hour fire barriers for the enclosure of cable and equipment and associated non-safety circuits of one redundant train (G.2.c). As used in this EAL, the 30-minutes to verify a single alarm is well within this worst-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 MAS.

CGS Basis Reference(s):

1. FSAR Table 3.2-1 Equipment Classification

2. NEI 99-01 HU4

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 115 of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 4-Fire Initiating Condition: FIRE potentially degrading the level of safety of the plant EAL:

HU4.3 Unusual Event (1) A FIRE within the ISFSI or plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)

OR (2) A FIRE within the ISFSI or plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

These thresholds reflect the potential issues that can arise from a fire in other areas of the plant for greater than one-hour or a fire requiring offsite fire department to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish.

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.

Threshold #1 In addition to a FIRE addressed by EAL HU4.1 or HU4.2, a FIRE within the plant PROTECTED AREA not extinguished within 60-minutes may also potentially degrade the level of plant safety. This basis extends to a FIRE occurring within the PROTECTED AREA of an ISFSI located outside the plant PROTECTED AREA.

Threshold #2 If a FIRE within the plant or ISFSI PROTECTED 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. The dispatch of an offsite firefighting agency to the site requires an emergency declaration only if it is needed to actively support firefighting efforts because the fire is beyond the capability of the Fire Brigade to extinguish. Declaration is not necessary if the agency resources are placed on stand-by, or supporting post-extinguishment recovery or investigation actions.

Depending upon the plant mode at the time of the event, escalation of the emergency classification level would be via IC CA6 or MAS.

CGS Basis Reference(s):

• 1. NEI 99-01 HU4 Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 116of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 5 - Hazardous Gases Initiating Condition: Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown EAL:

HA5.1 Alert Release of a toxic, corrosive, asphyxiant or flammable gas into any Table 9 rooms or areas AND Entry 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, then no emergency classification is warranted Table 9 Safe Operation & Shutdown Areas Room/Area Mode Applicability RW 467' Radwaste Control Room (RHR flush to RW tanks) 3

• RW 467' Vital Island (RHR-V-9 disconnect)

RB 422' B RHR Pump Rm (local pump temperatures)

RB 454' B RHR Pump Rm (operate RHR-V-85B) 3 3

3 Mode Applicability:

1 1 1 2 I 3 I 4 I s I def I Basis:

The list of plant rooms or areas with entry-related mode applicability identified 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 and shutdown. 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) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 1).

This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant 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 gaseous release. The emergency

_classification is not contingent upon whether entry is actually necessary at the time of the release.

Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1 A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 117of 190 Evaluation of the IC and EAL do not require atmospheric sampling; it only requires the Emergency Director's judgment that the gas concentration in the affected room/area is sufficient to preclude or significantly impede procedurally required access. This judgment may be based on a variety of factors including an existing job hazard analysis, 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 if 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).

An emergency declaration is not warranted if any of the following conditions apply:

• The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the gaseous 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 require entry into the affected room until Mode 4.

• The gas release is a planned activity that includes compensatory measures which address the 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 keeping nature (e.g., normal rounds or routine inspections).

• The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

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 can lead to breathing difficulties, unconsciousness or even death.

This EAL does not apply to firefighting activities that automatically or manually activate a fire suppression system in an area, or to intentional inerting of containment.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.

CGS Basis Reference(s):

1. Attachment 7.4 Safe Operation & Shutdown Areas Table 9 Bases

2. NEI 99-01 HAS

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 118 of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Inability to control a key safety function from outside the Control Room EAL:

HS6.1 Site Area Emergency An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel or Alternate Remote Shutdown Panel .

AND Control of any of the following key safety functions is not reestablished within 15 min. (Note 1):

• Reactivity (Modes 1 and 2 only)

• RPV water level

• RCS heat removal Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

• 1 Basis:

2 3 4 5 The Shift Manager determines if the Control Room is inoperable and requires evacuation. This determination can depend on a number of factors, including Control Room habitability, loss of safe -

shutdown control circuity, or a Security event (ref. 1).

For the purpose of this EAL the 15 minute classification clock starts when the last licensed operator leaves the Control Room.

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations, and the control of a key safety function cannot be reestablished in a timely manner. The failure to gain control of a key safety function following a transfer of plant control to alternate locations is a precursor to a challenge to one or more fission product barriers within a relatively short period of time.

The determination of whether or not "control" is established at the remote safe shutdown location(s) is based on Emergency Director judgment. The Emergency Director is expected to make a reasonable, informed judgment within 15 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 CG1.

CGS Basis Reference(s):

1. ABN-CR-EVAC Control Room evacuation and Remote Cooldown

2. NEI 99-01 HS6

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A J Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 119 of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 6 - Control Room Evacuation Initiating Condition: Control Room evacuation resulting in transfer of plant control to alternate locations EAL:

HA6.1 Alert An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel or Alternate Remote Shutdown Panel Mode Applicability:

1 1 1 2 1 3 1 4 1 5 1 def 1 Basis:

The Shift Manager (SM) determines if the Control Room is inoperable and requires evacuation. This determination can depend on a number of factors, including Control Room habitability, loss of safe shutdown control circuity, or a Security event (ref. 1). For the purpose of this EAL the 15 minute classification clock starts when the last licensed operator leaves the Control Room.

Inability to establish plant control from outside the Control Room escalates this event to a Site Area

• Emergency per EAL HS6.1 .

This IC addresses an evacuation of the Control Room that results in transfer of plant control to alternate locations outside the Control Room. The loss of the ability to control the plant from the Control Room is considered to be a potential substantial degradation in the level of plant safety.

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 Control Room, 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 ERO and emergency response facilities will assist in responding to these challenges.

Escalation of the emergency classification level woul.d be via IC HS6.

CGS Basis Reference(s):

1. ABN-CR-EVAC Control Room Evacuation and Remote Cooldown

2. NEI 99-01 HAS

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1:1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 120 of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of a General Emergency EAL:

HG7.1 General Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that results in an actual loss of physical control of the facility. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager(SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures. If required by the emergency classification or if deemed appropriate by the Emergency Director, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but 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 managing a major emergency.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a General Emergency.

CGS Basis Reference(s):

1. NEI 99-01 HG7

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 121 of 190 Category: H - Hazards and Othe.r Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions existing which in the judgment of the Emergency Director warrant declaration of a Site Area Emergency EAL:

HS7.1 Site Area Emergency Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or malicious acts, (1) toward site personnel or equipment that could lead to the likely failure of or, (2) that prevent effective access to equipment needed for the protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.

Mode Applicability:

1 1 1 2 1 3 1 4 1 5 1 def 1 Basis:

• The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures.

If required by the emergency classification or if deemed appropriate by the Emergency Director, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but 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 managing a major emergency.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for a Site Area Emergency.

CGS Basis Reference(s):

1. NEI 99-01 HS?

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 122 of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions exist which in the judgment of the Emergency Director warrant declaration of an Alert EAL:

HA7.1 Alert Other conditions exist which, in the judgment of the Emergency 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 involves probable life threatening risk to site personnel or damage to site equipment because of HOSTILE ACTION. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

I The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures .

If required by the emergency classification or if deemed appropriate by the Emergency Director, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but 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 managing a major emergency.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Alert.

CGS Basis Reference(s):

1. NEI 99-01 HA?

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 123 of 190 Category: H - Hazards and Other Conditions Affecting Plant Safety Subcategory: 7 - Emergency Director Judgment Initiating Condition: Other conditions existing which in the judgment of the Emergency Director warrant declaration of a UE EAL:

HU7.1 Unusual Event Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of SAFETY SYSTEMS occurs.

Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

The Emergency Director is the designated onsite individual having the responsibility and authority for implementing the Emergency Response Plan. The Shift Manager (SM) initially acts in the capacity of the Emergency Director and takes actions as outlined in the Emergency Plan implementing procedures .

If required by the emergency classification or if deemed appropriate by the Emergency Director, emergency response personnel are notified and instructed to report to their emergency response locations. In this manner, the individual usually in charge of activities in the Control Room is responsible for initiating the necessary emergency response, but 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 managing a major emergency.

This IC addresses unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the Emergency Director to fall under the emergency classification level description for an Unusual Event.

• CGS Basis Reference(s):

1. NEI 99-01 HU7

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Re_v: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 124 of 190 Category M - System Malfunction EAL Group: Hot Conditions (RCS temperature GT 200°F);

EALs in this category are applicable only in one or more hot operating modes.

Numerous system-related equipment failure events that warrant emergency classification have been 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 Emergency AC Power Loss of emergency electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of onsite and offsite sources for emergency AC buses.

2. Loss of vital DC Power Loss of vital electrical power can compromise plant safety system operability including decay heat removal and emergency core cooling systems which may be necessary to ensure fission product barrier integrity. This category includes loss of power to or degraded voltage on the vital 125 VDC buses.

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 this subcategory.

4. RCS Activity During normal operation, reactor coolant fission product activity is very low. Small concentrations of fission products in the coolant are primarily from the fission of tramp uranium in the fuel clad or minor perforations in the clad itself. Any significant increase from these base-line levels (2% - 5%

clad failures) is indicative of fuel failures and is covered under the Fission Product Barrier Degradation category. However, lesser amounts of 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 by coolant sampling.

5. RCS Leakage The reactor pressure vessel provides a volume for the coolant that covers the reactor core. The reactor pressure vessel and associated pressure piping (reactor coolant system) together provide a barrier to limit the release of radioactive material should the reactor fuel clad integrity fail. Excessive RCS leakage greater than Technical Specification limits indicates potential pipe cracks that may propagate to an extent threatening fuel clad, RCS and Containment integrity.

6. RPS Failure This subcategory includes events related to failure of the Reactor Protection System (RPS) to initiate and complete reactor scrams. In the plant licensing basis, postulated failures of the RPS to complete a reactor scram comprise a specific set of analyzed events referred to as Anticipated Transient Without Scram (ATWS) events. For EAL classification, however, A TWS is intended to mean any trip failure event that does not achieve reactor shutdown. If RPS actuation fails to assure reactor shutdown, positive control of reactivity is at risk and could cause a threat to fuel clad, RCS and Containment integrity.

Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 125 of 190

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. Hazardous Event Affecting Safety Systems Certain hazardous natural and technological events may result in visible damage to or degraded performance of safety systems warranting classification .

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifyi,ng The Emergency - Technical Bases Page: 126 of 190 Category: M -System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Prolonged loss of all offsite and all onsite AC power to emergency buses EAL:

MG1.1 General Emergency Loss of all offsite AND all onsite AC power capability to emergency buses SM-7 and SM-8 AND EITHER:

Restoration of emergency bus SM-7 or SM-8 in LT 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)

OR RPV level cannot be restored and maintained GT -186 in.

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

1 2 3 Basis:

• Credit may be taken in this EAL for DG 3 crosstie capability provided a reasonable expectation exists that AC power can be restored to either SM-7 or SM-8 from DG3 and SM-4 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. (ref. 1).

Four hours is the station blackout coping time (ref. 2).

Indication of continuing core cooling degradation is manifested by the inability to restore and maintain RPV water level above the Minimum Steam Cooling Reactor Water Level (-186 in.) (ref. 3).

Core submergence is the most desirable means of core cooling, however when RPV level is below TAF, the uncovered portion of the core can be cooled by less reliable means (i.e., steam cooling or spray cooling).

This IC addresses a prolonged loss of all power sources to AC emergency buses. A loss of all AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A prolonged loss of these buses will lead to a loss of one or more fission product barriers. In addition, fission product barrier monitoring capabilities may be degraded under these conditions.

The EAL should 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 is projected that power cannot be restored to at least one AC emergency bus by the end of the analyzed station blackout coping period. Beyond this time, plant responses and event trajectory are subject to greater uncertainty, and there is an increased likelihood of challenges to multiple fission product barriers.

The estimate for restoring at least one essential bus should be based on a realistic appraisal of the situation. Mitigation actions with a low probability of success should not be used as a basis for delaying a classification upgrade. The goal is to maximize the time available to prepare for, and implement, protective actions for the public.

Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 127 of 190 The EAL will also require a General Emergency declaration if the loss of AC power results in parameters that indicate an inability to adequately remove decay heat from the core.

CGS Basis Reference(s):

1. FSAR Section 8.2

2. PPM 5.6.1 Station Blackout (SBO)

3. PPM 5.1.1 RPV Control

4. NEI 99-01 SG1

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: .13.1.1 A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 128 of 190 Category: M - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite power and all onsite AC power to emergency buses for 15 minutes or longer EAL:

MS1.1 Site Area Emergency Loss of all offsite and all onsite AC power capability to emergency buses SM-7 and SM-8 for GE 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

1 2 3 Basis:

This* hot condition EAL is equivalent to the cold condition EAL CA2.1.

This IC addresses a total loss of AC power that compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink.

In addition, fission product barrier monitoring capabilities may be degraded under these conditions. This IC represents a condition that involves actual or likely major failures of plant functions needed for the protection of the public.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via ICs RG1, FG1 or MG1.

CGS Basis Reference(s):

1. PPM 5.6.1 Station Blackout (SBO)

2. NEI 99-01 SS1

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 129 of 190 Category: M - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all but one AC power source to emergency buses for 15 minutes or longer EAL:

MA1.1 Alert AC power capability, Table 2, to emergency buses SM-7 and SM-8 reduced to a single power source for GE 15 min. (Note 1)

AND Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 2 AC Power Sources I I Offsite Startup Transformer TR-S Backup Transformer TR-B Backfeed 500 KV power through Main Transformers (if already aligned in modes 4, 5, def only)

Onsite

• DG1

• DG2

• Main Generator via TR-N1/N2 Mode Applicability:

1 2 3 Basis:

Table 2 provides the list of AC power sources available to power emergency buses (ref. 1, 2).

Station Startup 230KV power comes from the Ashe substation through Startup transformer TR-S. The Startup transformer usually supplies station auxiliary loads when the main generator is not available.

Station Backup 115KV power from the Benton Substation feeder can be supplied to emergency buses SM-7 and SM-8 (ref. 3, 4) .

Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 130 of 190 Credit is not taken in this EAL for SM-4/DG3 crosstie capability because establishing the crosstie to SM-7 or SM-8 is assumed to require more than 15 minutes (5). SM-4 is not a site specific emergency AC buss source since SM-4 does not provide core cooling or containment cooling.

The second threshold statement in this EAL does not describe a separate condition, it is clarifying the first threshold statement.

This hot condition EAL is equivalent to the cold condition EAL CU2.1.

This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of safety-related equipment.

This IC provides an escalation path from IC MU1.

An "AC power source" is a source recognized in AOPs, and capable of supplying required power to an emergency bus. Some examples of this condition are presented below.

• A loss of all offsite power with a concurrent failure of all but one emergency power source (e.g., an onsite diesel generator).

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

• A loss of emergency power sources (e.g., onsite diesel generators) with a single train of

• emergency buses being back-fed from an offsite power source .

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of power.

Escalation of the emergency classification level would be via IC MS1.

CGS Basis Reference(s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. ABN-ELEC-LOOP

7. NEI 99-01 SA1

• Attachment 6.1, EAL Technical Bases

IUse Catego~: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 131 of 190 Category: M - System Malfunction Subcategory: 1 - Loss of Emergency AC Power Initiating Condition: Loss of all offsite AC power capability to emergency buses for 15 minutes or longer EAL:

MU1.1 Unusual Event Loss of all offsite AC power capability, Table 2, to emergency buses SM-7 and SM-8 for GE 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 2 AC Power Sources I I Offsite

• Startup Transformer TR-S

• Backup Transformer TR-B

• Backfeed 500 KV power through Main Transformers (if already aligned in modes 4, 5, def only)

Onsite

• DG1

• DG2

• Main Generator via TR-N1/N2 Mode Applicability:

1 2 3 Basis:

Table 2 provides the list of AC power sources available to power emergency buses (ref. 1, 2).

Station Startup 230KV power comes from the Ashe substation through Startup transformer TR-S. The Startup transformer usually supplies station auxiliary loads when the main generator is not available.

Station Backup 115KV power from the Benton Substation feeder can be supplied to emergency buses SM-7 and SM-8. (ref. 3, 4)

Credit is not taken in this EAL for SM-4/DG3 crosstie capability because establishing the crosstie to SM-7 or SM-8 is assumed to require more than 15 minutes (5). SM-4 is not a site specific emergency AC buss source since SM-4 does not provide core cooling or containment cooling .

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A J Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 132 of 190 This IC addresses a prolonged loss of offsite power. The loss of offsite power sources renders the plant more vulnerable to a complete loss of power to AC emergency 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 emergency buses, whether or not the buses are powered from it.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of offsite power.

Escalation of the emergency classification level would be via IC MA 1.

CGS Basis Reference(s):

1. FSAR Figure 8.1-2.1 Main One-Line Diagram - Main Buses

2. FSAR Figure 8.1-2.2 Main One-Line Diagram - Emergency Buses

3. FSAR Section 8.2

4. 01-53 Offsite Power

5. FSAR Section 8.3

6. NEI 99-01 SU1

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 133 of 190 Category: M -System Malfunction Subcategory: 1 - Loss of Essential AC Power

• Initiating Condition: Loss of all emergency AC and vital DC power sources for 15 minutes or longer EAL:

MG1 .2 General Emergency Loss of all offsite AND all onsite AC power capability to emergency buses SM-7 and SM-8 for GE 15 min. (Note 1)

AND Indicated voltage is LT 108 VDC on both 125 voe buses DP-S1-1 and DP-S1-2 for GE 15 min.

(Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

1 2 3 Basis:

• This EAL addresses operating experience from the March 2011 accident at Fukushima Daiichi.

The 125 VDC Class 1E DC power system consists of three electrically independent and separate distribution systems (S1-1, S1-2, and S1-HPCS) (ref. 2). S1-HPCS is not included in this EAL. Each DC distribution system has a battery and a battery charger that are normally connected to the bus such that these two sources of power are operating in parallel. The charger is normally supplying system electrical loads with the battery on a float charge. Each battery has the necessary amp-hour discharge capacity to sustain system loads for a minimum of two hours. This capacity is specifically for a loss of power to the charger coincident with a design basis accident. The batteries have capacity to carry design load at 60°F without decreasing battery voltage below 1.81 volts/cell (or 108 VDC, ref. 3) with loss of output from the battery chargers during the specified period. Battery capacity is sufficient to provide starting currents while operating at full load. (ref. 1, 3). '

This IC addresses a concurrent and prolonged loss of both emergency AC and Vital DC power. A loss of all emergency AC power compromises the performance of all SAFETY SYSTEMS requiring electric power including those necessary for emergency core cooling, containment heat removal/pressure control, spent fuel heat removal and the ultimate heat sink. A loss of vital DC power compromises the ability to monitor and control SAFETY SYSTEMS. A sustained loss of both emergency AC and vital DC power will lead to multiple challenges to fission product barriers.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. The 15-minute emergency declaration clock begins at the point when both EAL thresholds are met.

CGS Basis Reference(s):

1. FSAR Section 8

2. E505 DC One Line Diagram

3. Calculation No. 2.05.01 Battery Sizing, Voltage Drop, and Charger Studies for Div. 1 & 2 Systems

4. PPM 5.6.1 Station Blackout (SBO)

5. NEI 99-01 SG8 Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 134 of 190 Category: M - System Malfunction Subcategory: 2 - Loss of Vital DC Power Initiating Condition: Loss of all vital DC power for 15 minutes or longer EAL:

MS2.1 Site Area Emergency Indicated voltage is LT 108 voe on both 125 voe buses DP-S1-1 and DP-S1-2 for GE 15 min.

(Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

1 2 3 Basis:

The 125 VDC Class 1E DC power system (ref. 1) consists of three electrically independent and separate distribution systems (S1-1, S1-2, and S1-HPCS) (ref.-2). S1-HPCS is not included in this EAL.

Each DC distribution system has a battery and a battery charger that are normally connected to the bus such that these two sources of power are operating in parallel. The charger is normally supplying system electrical loads with the battery on a float charge. Each battery has the necessary amp-hour discharge capacity to sustain system loads for a minimum of two hours. This capacity is specifically for a loss of power to the charger coincident with a design basis accident. The batteries have capacity to carry design load at 60°F without decreasing battery voltage below 1.81 volts/cell (or 108 VDC, ref. 2) with loss of output from the battery chargers during the specified period. Battery capacity is sufficient to provide starting currents while operating at full load. (ref. 3).

This EAL is the hot condition equivalent of the cold condition loss of DC power EAL CU4.1.

This IC addresses a loss of vital DC power which compromises the ability to monitor and control SAFETY SYSTEMS. In modes above Cold Shutdown, this condition involves a major failure of plant functions needed for the protection of the public.

Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation of the emergency classification level would be via ICs RG1, FG1 or MG1.

CGS Basis Reference(s):

1. E505 DC One Line Diagram

2. Calculation No. 2.05.01 Battery Sizing; Voltage Drop, and Charger Studies for Div. 1 & 2 Systems

3. FSAR Section 8.3.2

4. NEI 99-01 SS8

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A \

Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 135 of 190 Category: M - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress EAL:

MA3.1 Alert An UNPLANNED event results in the inability to monitor one or more Table 1O parameters from within the Control Room for GE 15 min. (Note 1)

AND Any Table 11 transient event in progress Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 10 Safety System Parameter:;.... 11

.. Reactor power RPV level

• RPV pressure

• Primary containment pressure

• Wetwell level

• Wetwell temperature Table 11 Significant Transients

• Reactor scram

• Runback GT 25% thermal reactor power

• Electrical load rejection GT 25% full electrical load

• ECCS injection

• Thermal power oscillations GT 10%

Mode Applicability:

2 3

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency- Technical Bases Page: 136 of 190 Basis:

SAFETY SYSTEM parameters listed in Table 1O are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems. The Plant Computers and Graphic Display System provide redundant parameter indications (ref. 1-4).

Significant transients are listed in Table 11 and include response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, electrical load rejections of greater than 25% full electrical load, ECCS injections, or thermal power oscillations of 10%

or greater.

This IC addresses the difficulty associated. with monitoring rapidly changing plant conditions during a transient withou~ the ability to obtain SAFETY SYSTEM parameters from within the Control Room.

During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for on'3 or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRG event report is required. The event would be reported if it significantly impaired the capability to perform emergency assessments. In particular, emergency assessments necessary to implement abnormal operating procedures, emergency operating procedures, and emergency plan implementing procedures addressing emergency classification, accident assessment, or protective action decision-making.

This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more

' of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via ICs FS1 or IC RS1 CGS Basis Reference(s):

1. FSAR Section 7.7.1

2. ABN-COMPUTER

3. SOP-COMPUTER-OPS Plant Process Computer (PPG)

4. SOP-GOS-OPS Graphics Display System

5. NEI 99-01 SA2

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 137 of 190 Category: M - System Malfunction Subcategory: 3 - Loss of Control Room Indications Initiating Condition: UNPLANNED loss of Control Room indications for 15 minutes or longer EAL:

MU3.1 Unusual Event An UNPLANNED event results in the inability to monitor one or more Table 10 parameters from within the Control Room for GE 15 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Table 10 Safety System Parameters

• Reactor power

• RPV level

• RPV pressure

• Mode Applicability:

Primary containment pressure Wetwell level Wetwell temperature 1 2 3 Basis:

SAFETY SYSTEM parameters listed in Table 10 are monitored in the Control Room through a combination of hard control panel indicators as well as computer based information systems. The Plant Computers and Graphic Display System provide redundant parameter indications (ref. 1-4).

This IC addresses the difficulty associated with monitoring normal plant conditions without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. This condition is a precursor to a more significant event and represents a potential degradation in the level of safety of the plant.

As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

An event involving a loss of plant indications, annunciators and/or display systems is evaluated in accordance with 10 CFR 50.72 (and associated guidance in NUREG-1022) to determine if an NRC event report is required. The event 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.

Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 138 of 190 This EAL is focused on a selected subset of plant parameters associated with the key safety functions of reactivity control, RPV level and RCS heat removal. The loss of the ability to determine one or more of these parameters from within the Control Room is considered to be more significant than simply a reportable condition. In addition, if all indication sources for one or more of the listed parameters are .

lost, then the ability to determine the values of other SAFETY SYSTEM parameters may be impacted as well. For example, if the value for RPV water level cannot be determined from the indications and recorders on a main control board, the SPDS or the plant computer, the availability of other parameter values may be compromised as well.

Fifteen minutes was selected as a threshold to exclude transient or momentary losses of indication.

Escalation of the emergency classification level would be via IC MA3.

CGS Basis Reference(s):

1. FSAR Section 7.7.1

2. ABN-COMPUTER

3. SOP-COMPUTER-OPS Plant Process Computer (PPC)

4. SOP-GOS-OPS Graphics Display System

5. NEI 99-01 SU2

• Attachment 6.1, EAL Technical Bases

\

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 139 of 190 Category: M - System Malfunction Subcategory:

• 4 - RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:

MU4.1 Unusual Event SJAE CONDSR OUTLET RAD HI-HI alarm (P602)

Mode Applicability:

1 2 3 Basis:

The main condenser offgas gross gamma activity rate is an initial condition of the Main Condenser Offgas System failure event. The gross gamma activity rate is controlled to ensure that during the event, the calculated offsite doses will be well within the limits of 10 CFR 50.67 (ref. 1).

SJAE CONDSR OUTLET RAD HI HI monitor and alarm, OG-RIS-612 (GE 2300 mR/hr), senses the offgas effluent and, therefore, may be one of the first indicators of degrading fuel conditions. The alarm is confirmed by verification of greater than the current alarm setpoint on Recorder OG-RIS-612 on Panel P604 or high offgas pre-treatment air activity (determined by sample results) greater than limits

• specified in Technical Specification .

If OG-RIS-612 and OG-RR-604 are reading off-scale high, the alarm may be confirmed by a significant increase in the Main Steam Line radiation monitors (MS-RIS-610A-D) on H13-P606 and H13-P633 (ref. 2).

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via ICs FA1 or the Recognition Category R ICs.

CGS Basis Reference(s):

1. Technical Specifications 3. 7 .5

2. PPM 4.602.AS ANNUNCIATOR RESPONSE, P602 ANNUNCIATOR AS 3-3

3. NEI 99-01 SU3

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 140 of 190 Category: M - System Malfunction Subcategory: 4 - RCS Activity Initiating Condition: Reactor coolant activity greater than Technical Specification allowable limits EAL:

MU4.2 Unusual Event Coolant activity GT 0.2 µCi/gm dose equivalent 1-131 Mode Applicability:

1 2 3 Basis:

The limits on the specific activity of the primary coolant ensure that the 2-hour thyroid and whole body doses at the SITE BOUNDARY, resulting from an Main Steam Line Break (MSLB) outside containment during steady state operation, will not exceed the dose guidelines of 10 CFR 50.67.

This IC addresses a reactor coolant activity value that exceeds an allowable limit specified in Technical Specifications. This condition is a precursor to a more significant event and represents a potential degradation of the level of safety of the plant.

Escalation of the emergency classification level would be via ICs FA 1 or the Recognition Category R ICs.

CGS Basis Reference(s):

1. Technical Specifications 3.4.8

2. NEI 99-01 SU3

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency.- Technical Bases Page: 141 of 190 Category: M - System Malfunction Subcategory: 5 - RCS Leakage Initiating Condition: RCS leakage for 15 minutes or longer EAL:

MU5.1 Unusual Event (1) RCS unidentified or pressure boundary leakage GE 10 gpm for GE 15 min.

OR (2) RCS identified leakage GT 25 gpm for GE 15 min.

OR (3) Leakage from the RCS to a location outside containment GT 25 gpm for GE 15 min.

(Note 1)

Note 1: The Emergency Director sh'ould declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

1 2 3

• Basis:

Pressure boundary leakage is defined to be leakage through a non-isolable fault in a RCS component body, pipe wall, or vessel wall.

This EAL does not apply to relief valves performing their normal design function.

Unidentified leakage is defined to be all leakage into the drywell that is not identified leakage.

Identified leakage is defined to be leakage into the drywell such as that from pump seals or valve packing, that is captured and conducted to a sump; or leakage into the drywell atmosphere from sources 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. (ref. 1)

The Leak Detection (LD) system is designed to monitor leakage from the reactor coolant pressure boundary and to isolate this leakage when limits are exceeded. Systems or parts of systems that are in direct communication with the reactor vessel (form part of the primary coolant pressure boundary) are provided with leakage detection systems. (ref. 2-8)

Drain flow from the drywell equipment and floor drain sumps is monitored and recorded (EDR-FRS-623) on P632. The flow rate for unidentified leakage in the EAL is equal to the full scale reading on EDR-FRS-623 Pen 1.

Leakage not explicitly identified by installed instrumentation requires analysis and declaration clock starts*at completion of analysis. This includes use of alternate means.

As an alternate means, leaks within the drywell are detected by monitoring for abnormally high:

• Pressure or temperature inside the drywell

• Fill up rates of equipment and floor drain sumps

• Containment leak detection rad monitors (CMS-SR-20/21)

Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 142 of 190 Outside Containment leakage may require analysis to quantify leak rate GT 25 gpm and declaration clock starts at completion of analysis.

Examples of outside Containment leakage include:

• GT 25 gpm RWCU differential flow (LD-Fl-620) due to RCS leakage

• Instrument line break in the RX building with failure to isolate

• Rx Building sump fill timers due to RCS leakage RFW and RCC are not considered part of RCS leakage for this EAL.

For classification under this EAL, RCS leakage includes a broken SRV tailpipe that is discharging into the drywell or wetwell airspace. Once the SRV is closed, however, this RCS leakage path is considered isolated.

This IC addresses RCS leakage which may be a precursor to a more significant event. In this case, RCS leakage has been detected and operators, following applicable procedures, have been unable to promptly isolate the leak. This condition is considered to be a potential degradation of the level of safety of the plant.

Threshold #1 and threshold #2 are focused on a loss of mass from the RCS due to "unidentified leakage", "pressure boundary leakage" or "identified leakage" (as these leakage types are defined in the plant Technical Specifications). Threshold #3 addresses an RCS mass loss caused by an UNISOLABLE leak through an interfacing system. These conditions thus apply to leakage into the

. containment, or a location outside of containment.

The leak rate values for each threshold were selected because they are usually observable with normal Control Room indications. Lesser values typically require time-consuming calculations to determine (e.g., a mass balance calculation). Threshold #1 uses a lower value that reflects the greater significance of unidentified or pressure boundary leakage.

A stuck-open Safety Relief Valve (SRV) or SRV leakage is not considered either identified or unidentified leakage by Technical Specifications and, therefore, is not applicable to this EAL.

The 15-minute threshold duration allows sufficient time for prompt operator actions to isolate the leakage, if possible.

Escalation of the emergency classification level would be via ICs of Recognition Category R or F.

CGS Basis Reference(s):

1. Technical Specification 1.1

2. Technical Specifications 3.4. 7

3. FSAR Section 5.2.5

4. FSAR Section 7.6.1

5. ABN-LEAKAGE Reactor Coolant Leakage

6. SOP-EDR-OPS Equipment Drain System Operation

7. SOP-FDR-OPS Floor Drain System Operation

8. PPM 10.27.35 Leakage Surveillance And Prevention Program

9. NEI 99-01 SU4

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 143 of 190 Category: M - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Inability to shut down the reactor causing a challenge to RPV water level or RCS heat removal EAL:

MS6.1 Site Area Emergency An automatic OR manual scram fails to shut down the reactor AND All actions to shut down the reactor are not successful as indicated by reactor power GT 5%

AND EITHER:

RPV level cannot be restored and maintained above -186 in. or cannot be determined OR WW temperature and RPV pressure cannot be maintained below the HCTL Mode Applicability:

1 2 Basis:

This EAL addresses the following:

• Any automatic reactor scram signal followed by a manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed (EAL MA6.1), and

• Indications that either core cooling is extremely challenged or heat removal is extremely challenged.

Reactor shutdown achieved by use of control rod insertion methods in PPM 5.5.11 is also credited as a successful manual scram provided reactor power can be reduced below the APRM downscale trip setpoint before indications of an extreme challenge to either core cooling or heat removal exist. (ref. 1)

I The APRM downscale trip setpoint (5%) is a minimum reading on the power range scale that indicates power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM) indications or other reactor parameters (steam flow, RPV pressure, wetwell temperature trend) can be used to determine if reactor power is greater than 5% power (ref. 2).

The combination of failure of both front line and backup protection systems to function in response to a plant transient, along with the continued production of heat, poses a direct threat to the Fuel Clad and RCS barriers .

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 144 of 190 Indication that core cooling is extremely challenged is manifested by inability to restore and maintain RPV water level above the Minimum Steam Cooling RPV Water Level (MSCRWL) (ref. 3). The MSCRWL is the lowest RPV level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500°F. This water level is utilized in the EOPs to preclude fuel damage when RPV level is below the top of active fuel. RPV level below the MSCRWL for an extended period of time without satisfactory core spray cooling could be a precursor of a core melt sequence.

The Heat Capacity Temperature Limit (HCTL) is the highest suppression pool water temperature from which Emergency RPV Depressurization will not raise suppression pool temperature above the maximum design suppression pool temperature.

The HCTL is a function of RPV pressure and wetwell level. It is utilized to preclude failure of the containment and equipment in the containment necessary for the safe shutdown of the plant (ref. 4).

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances, the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F ICs/EALs. This is appropriate in that the Recognition Category F ICs/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 in response to prolonged failure to shutdown the reactor.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Escalation of the emergency classification level would be via IC RG1 or FG1.

CGS Basis Reference(s):

1. PPM 5.5.11 Alternate Control Rod Insertions

2. Technical Specifications Table 3.3.1.1-1

3. PPM 5.1.2 RPV Control - ATWS

4. PPM 5.2.1 Primary Containment Control,

5. NEI 99-01 SS5

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1 A Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 145 of 190 Category: M - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor and subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor EAL:

MA6.1 Alert An automatic OR manual scram fails to shut down the reactor AND Manual scram actions taken at the reactor control console (mode switch in shutdown, manual push buttons or ARI) are not successful in shutting down the reactor as indicated by reactor power GT 5% (Note 8)

Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies Mode Applicability:

1 2 Basis:

This EAL addresses any automatic or manual reactor scram signal that fails to shut down the reactor followed by a subsequent manual scram that fails to shut down the reactor to an extent the reactor is producing energy in excess of the heat load for which the safety systems were designed.

For the purposes of emergency classification, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., mode switch in shutdown, manual push buttons or ARI). Reactor shutdown achieved by use of the alternate control rod insertion methods of PPM 5.5.11 does not constitute a successful manual scram (ref. 1).

The APRM downscale trip setpoint (5%) is a minimum reading on the power range scale that indicates power production. It also approximates the decay heat which the shutdown systems were designed to remove and is indicative of a condition requiring immediate response to prevent subsequent core damage. Below the APRM downscale trip setpoint, plant response will be similar to that observed during a normal shutdown. Nuclear instrumentation (APRM) indications or other reactor parameters (steam flow, BPV position or continuous SRV operation) can be used to determine if reactor power is greater than 5% power (ref. 2).

Escalation of this event is via EAL MS6.1.

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and subsequent operator manual actio_ns taken at the reactor control consoles to shutdown the reactor are also unsuccessful. This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even if the reactor is subsequently shutdown by an action taken away from the reactor control consoles since this event entails a significant failure of the RPS .

Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 146 of 190 A manual action at the reactor control console is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g.; initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies. If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the reactor control console (e.g., locally opening breakers). Actions taken at backpanels or other locations within the control room, or any location outside the control room, are not considered to be "at the reactor control console".

Taking the reactor mode switch to shutdown is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shut down the reactor is prolonged enough to cause a challenge to RPV water level or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC MS6.

Depending upon plant responses and symptoms, escalation is also possible via IC FS1. Absent the plant conditions needed to meet either IC MS6 or FS1, an Alert declaration is appropriate for this event.

It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

• CGS Basis Reference(s):

1. PPM 5.5.11 Alternate Control Rod Insertions

2. Technical Specifications Table 3.3.1.1-1

3. NEI 99-01 SAS

• Attachment 6.1, EAL Technical Bases

I Use Category:

Number: 13.1.1A REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 147 of 190 Category: M - System Malfunction Subcategory: 6 - RPS Failure Initiating Condition: Automatic or manual scram fails to shut down the reactor EAL:

MU6.1 Unusual Event An automatic OR manual scram did not shut down the reactor AND A subsequent automatic scram OR manual scram action taken at the reactor control console (mode switch in shutdown, manual push buttons or ARI) is successful in shutting down the reactor as indicated by reactor power LE 5% (APRM downscale) (Note 8)

Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies Mode Applicability:

1 2 Basis:

This EAL addresses a failure of an automatic or manually initiated scram and a subsequent automatic or manual scram is successful in shutting down the reactor (reactor power LE 5%) (ref.1).

A successful scram has occurred when there is sufficient rod insertion from the trip of RPS to bring the reactor power to or below the APRM downscale trip setpoint of 5%. For the purposes of this EAL, a successful automatic initiation of ARI that reduces reactor power to or below 5% is a not a successful automatic scram. (ref. 2, 3, 4, 5)

For the purposes of emergency classification at the Unusual Event level, successful manual scram actions are those which can be quickly performed from the reactor control console (i.e., mode switch, manual scram pushbuttons, and manual ARI actuation). Reactor shutdown achieved by use of the alternate control rod insertion methods of PPM 5.5.11 does not constitute a successful manual scram (ref. 6).

Following any automatic RPS scram signal plant procedures prescribe insertion of redundant manual scram signals to back up the automatic RPS scram function and ensure reactor shutdown is achieved.

Even if the first subsequent manual scram signal inserts all control rods to the full-in position immediately after the initial failure of the automatic scram, the lowest level of classification that must be declared is an Unusual Event.

The operating mode change associated with movement of the Mode Switch, by itself, does not justify failure to declare an emergency for A TWS events.

If both subsequent automatic and subsequent manual reactor scram actions in the Control Room fail, the event escalates to an Alert under EAL MA6.1.

This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor scram that results in a reactor shutdown, and either a subsequent operator manual action taken at the reactor control consoles or an automatic scram is successful in shutting down the reactor. This event is a Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical.Bases *

• Page: 148 of 190 precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant.

• Following the failure on an automatic reactor scram, operators will promptly initiate manual actions at the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram). If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

If an initial manual reactor scram is unsuccessful, operators will promptly take manual action at another location(s) on the reactor control consoles to shutdown the reactor (e.g., initiate a manual reactor scram1 using a different switch). Depending upon several factors, the initial or subsequent effort to manually scram the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor scram signal. If a subsequent manual or automatic scram is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems.

A manual action at the reactor control consoles is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core (e.g., initiating a manual reactor scram). This action does not include manually driving in control rods or implementation of boron injection strategies.

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

Taking the reactor mode switch to shutdown is considered to be a manual scram action.

The plant response to the failure of an automatic or manual reactor scram will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the reactor control consoles are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an alert via IC MAS. Depending upon the plant response, escalation is also possible via IC FA 1. Absent the plant conditions needed to meet either IC MAS or FA 1, an unusual event declaration is appropriate for this event.

A reactor shutdown is determined in accordance with applicable emergency operating procedure criteria.

Should a reactor scram signal be generated as a result of plant work (e.g., RPS setpoint testing), the following classification guidance should be applied.

• If the signal causes a plant transient that should have included an automatic reactor scram and the RPS fails to automatically shutdown the reactor, then this IC and the EALs are applicable, _

and should be evaluated.

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

CGS Basis Reference(s):

1. Technical Specifications Table 3.3.1.1-1

2. FSAR Section 7.2

3. FSAR Section 7.4

4. PPM 5.1.1 RPV Control

5. PPM 5.1.2 RPV Control-ATWS Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 149 of 190

6. PPM 5.5.11 Alternate Control Rod Insertions

7. NEI 99-01 SUS

• Attachment 6.1, EAL Technical Bases

I Use Category:

Number: 13.1.1A REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 150 of 190 Category: M - System Malfunction Subcategory: 7 - Loss of Communications Initiating Condition: Loss of~ onsite or offsite communications capabilities EAL:

MU7.1 Unusual Event (1) Loss of~ Table 4 onsite communication methods OR (2) Loss of~ Table 4 ORO communication methods OR (3) Loss of~ Table 4 NRC communication methods Table 4 Communication Methods System Onsite ORO NRC

• Plant Public Address (PA) System Plant Telephone System Plant Radio System Operations and Security Channels X

X X

X Offsite calling capability from the Control X X Room via direct telephone Long distance calling capability on the X X commercial phone system Mode Applicability:

1 2 3 Basis:

Onsite and offsite (ORO and NRC) communications include one or more of the systems listed in Table 4 (ref. 1, 2).

Public Address (PA) System The public address system provides a way of contacting personnel in the various buildings of the plant and locations of the site that might be inaccessible using other means of communication. The building-wide alarm system alerts (via the public address system speakers) operating personnel to fire hazards

• and other trouble conditions for which plant management finds it necessary to alert plant personnel.

Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 151 of 190 Plant Telephone System This system consists of interconnections to the public telephone network (and trunks to the PBX) with individual direct lines that provide inward and outward dialing access to most plant locations.

Plant Radio System Operations and Security Channels The radio communications system is used for communications with personnel involved in maintenance and security in and around the plant complex by means of hand-held portable radio units, mobile radio units, and paging receivers. The telephone link to BPA provides a direct communication link to the BPA Dittmer Control Center. The radio communications system provides a communications link for security and emergency communications to local law enforcement agencies and emergency control centers.

Offsite calling capability from the Control Room via direct telephone and fax lines This communications method includes following dedicated phone networks that are available for emergency communications in addition to the normal Energy Northwest phone network:

• Energy Northwest Emergency Center Network

• Response Agency Network

• NRC Emergency Notification System Various locations such as the Control Room, Technical Support Center, Emergency Operations Facility, Joint Information Center, Department of Energy-RL, Washington State Emergency Operations Center, Oregon State Emergency Coordination Center and the Benton and Franklin County Emergency Operations Centers have facsimile transceivers. The facsimile transceivers enable the transmission and receipt of printed material. The facsimile system which connects the Energy Northwest emergency centers with the county and state emergency centers uses dedicated phone lines.

Long distance calling capability on the commercial phone system The Energy Northwest Richland phone system is a computer based, software controlled telephone exchange (Computerized Branch Exchange). It is equipped with redundant computerized processor units and is served by an uninterruptible power supply. The direct-dial private telephone system provides communication between the Energy Northwest facilities. The phone system is arranged such that plant telephones can reach other Energy Northwest facilities by direct-dialing and without the need of an operator.

• This EAL is the hot condition equivalent of the cold condition EAL CU5.1.

This IC addresses a significant loss of on-site or offsite communications capabilities. While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC.

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

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

Threshold #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaratiori. The OROs referred to here are Washington State, Benton County, Franklin County and DOE RL.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A \ Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 152 of 190 Threshold #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

CGS Basis Reference(s):

1. Emergency Plan Section 6.6

2. FSAR Section 9.5.2

3. NEI 99-01 SU6

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 153 of 190 Category: M - System Malfunction Subcategory: 8 - Hazardous Event Affecting Safety Systems Initiating Condition: Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode EAL:

MAS.1 Alert The occurrence of any Table 8 hazardous event AND Event damage has caused indications of degraded performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

Event damage has caused indications of degraded performance to a second train of a SAFETY SYSTEM needed for the current operating mode OR Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating mode

• Note 9:

(Notes 9, 10)

If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 10: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted.

Table 8 Hazardous Events

• Seismic event

• Internal or external FLOODING event

• High winds

• Tornado strike

• FIRE

• EXPLOSION

• Volcanic ash fallout

• Other events with similar hazard characteristics as determined by the Shift Manager

• Mode Applicability:

Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 154 of 190 1 2 3 Basis:

This IC addresses a hazardous event that causes damage to SAFETY SYSTEMS needed for the current operating mode. In order to provide the appropriate context for consideration of an ALERT classification, the hazardous event must have caused indications of degraded SAFETY SYSTEM performance in one train, and there must be either indications of performance issues with the second SAFETY SYSTEM train or VISIBLE DAMAGE to the second train such that the potential exists for this second SAFETY SYSTEM train to have performance issues. In other words, in order for this EAL to be classified, the hazardous event must occur, at least one SAFETY SYSTEM train must have indications of degraded performance, and the second SAFETY SYSTEM train must have indications of degraded performance or VISIBLE DAMAGE such that the potential exists for performance issues. Note that this second SAFETY SYSTEM train is from the same SAFETY SYSTEM that has indications of degraded performance; commercial nuclear power plants are designed to be able to support single system issues without compromising public health and safety from radiological events.

Indications of degraded performance addresses damage to a SAFETY SYSTEM train that is in service/operation since indications for it will be readily available. The indications of degraded performance should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

VISIBLE DAMAGE addresses damage to a SAFETY SYSTEM train that is not in service/operation and that potentially could cause performance issues. Operators will make this determination based on the totality of available event and damage report information. This is intended to be a brief assessment not requiring lengthy analysis or quantification of the damage. This VISIBLE DAMAGE should be significant enough to cause concern regarding the operability or reliability of the SAFETY SYSTEM train.

The significance of a seismic event is 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).

Seismic Category I structures are analyzed to withstand a sustained, design wind velocity of at least 100 mph (ref. 3).

Areas containing functions and systems required for safe shutdown of the plant are identified by Fire Areas in the fire response procedure (ref. 4).

The potential for volcanic eruption exists in the Pacific Northwest. Heavy ash fall, such as that experienced at certain locations following the eruption of Mt. St. Helens in 1980, could affect operation of plant equipment if precautionary measures are not taken. The design basis ash fall is projected for a twenty hour duration (ref. 5).

Table 5 provides a list of CGS safety system structures/areas (ref. 6). Table 8 provides a list of hazardous events.

Escalation of the emergency classification level would be via IC FS1 or RS1 .

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033

Title:

Classifying The Emergency - Technical Bases Minor Rev: N/A Page: 155 of 190 CGS Basis Reference(s):

1. FSAR Section 3. 7 Seismic Design

2. FSAR Section 3.4.1 Flood Protection

3. CGS Calculation CALC CE-02-93-16 Evaluate PMR/BDC 98-0131-0A change from 5 min. to 15 min. averaging of 33 ft. elev. met twr. wind speeds for UE and Alert declarations

4. ABN-FIRE Attachment 13.2, Fire Areas

5. ABN-ASH Ash Fall

6. FSAR Table 3.2-1 Equipment Classification

7. NEI 99-01 SA9

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 156 of 190 Category R - Abnormal Rad Release / Rad Effluent EAL Group: ANY (EALs in this category are applicable to any plant condition, hot or cold.)

Many EALs are based on actual or potential degradation of fission product barriers because of the elevated potential for offsite radioactivity release. Degradation of fission product barriers though is not always apparent via non-radiological symptoms. Therefore, direct indication of elevated radiological effluents or area radiation levels are appropriate symptoms for emergency classification.

At lower levels, abnormal radioactivity releases may be indicative of a failure of containment systems or precursors to more significant releases. At higher release rates, offsite radiological conditions may result which require offsite protective actions. Elevated area radiation levels in the plant may also be indicative of the failure of containment systems or preclude access to plant vital equipment necessary to ensure plant safety.

Events of this category pertain to the following subcategories:

1. Radiological Effluent Direct indication of effluent radiation monitoring systems provides a rapid assessment mechanism to determine releases in excess of classifiable limits. Projected offsite doses, actual offsite field measurements or measured release rates via sampling indicate doses or dose rates above classifiable limits.

• 2. Irradiated Fuel Event Conditions indicative of a loss of adequate shielding or damage to irradiated fuel may preclude access to vital plant areas or result in radiological releases that warrant emergency classification.

3. Area Radiation Levels Sustained general area radiation levels which may preclude access to areas requiring continuous occupancy also warrant emergency classification .

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 157 of 190 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEDE or 5,000 mrem thyroid COE EAL:

RG1 .1 General Emergency (1) Reading on any Table 3 effluent radiation monitor GT column "GENERAL" for GE 15 min.

OR (2) Dose assessment using actual meteorology indicates doses GT 1,000 mrem TEDE or GT 5,000 mrem thyroid COE at or beyond the SIJE BOUNDARY (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit

• Note 3:

Note 4:

If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes The pre-calculated effluent monitor values presented in EALs RA 1.1, RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Table 3 Effluent Monitor Classification Thresholds Release Point Monitor General SAE Alert UE PRM-RE-11 ---- ---- ---- 3.0SE-03 µCi/cc Reactor Building Exhaust PRM-RE-12 ---- ---- ' 2.82E+1 µCi/cc ----

Ill

l 0

Q) PRM-RE-13 7.50E+2 µCi/cc 7.50E+1 µCi/cc ---- ----

Ill C'G C) Turbine Building Exhaust TEA-RIS-13 8.35E-02 µCi/cc 8.35E-03 µCi/cc 8.35E-04 µCi/cc 4.22E-05 µCi/cc Radwaste Building WEA-RIS-14 3.45E-01 µCi/cc 3.45E-02 µCi/cc 3.45E-03 µCi/cc 3.98E-04 µCi/cc Exhaust Radwaste Effluent FDR-RIS-606 ---- ---- ---- 2 X HI-HI alarm "C TSWEffluent TSW-RIS-5 ---- ---- ---- 3.00E-05 µCi/cc

l C"

J Service Water Process A SW-RIS-604 ---- ---- ---- 1.00E+02 cps Service Water Process B SW-RIS-605 ---- --- --- 1.00E+02 cps Attachment 6.1, EAL Technical Bases

Number: 13.1.1A Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 158 of 190 Mode Applicability:

1 1 1 2 1 3 4 s I def 1 Basis:

Threshold #1 The pre-calculated effluent monitor values presented in Table 3 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to SITE BOUNDARY doses that exceed either (ref. 1, 2):

• 1000 mRem TEDE

• 5000 mRem COE Thyroid The column "GENERAL" gaseous effluent release values in Table 3 correspond to calculated doses of 100% of the EPA Protective Action Guidelines (TEDE or COE Thyroid) (ref. 1).

Threshold #2 Dose assessments are performed by computer-based methods (ref. 3) .

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

Classification based on effluent monitor readings assumes that a release path to the environment is ,

established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

CGS Basis Reference(s): *

1. Calculation NE-02-09-12 Revision 3

2. FSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling

3. PPM 13.8.1 Emergency Dose Projection System Operations

4. NEI 99-01 AG1

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 159 of 190 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL:

RS1.1 Site Area Emergency (1) Reading on any Table 3 effluent radiation monitor GT column "SAE" for GE 15 min.

OR (2) Dose assessment using actual meteorology indicates doses GT 100 mrem TEDE or GT 500 mrem thyroid COE at or beyond the SITE BOUNDARY (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes Note 4: The pre-calculated effluent monitor values presented in EALs RA 1.1, RS 1.1 and RG 1.1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Table 3 Effluent Monitor Classification Thresholds Release Point Monitor General SAE Alert UE PRM-RE-11 ---- ---- ---- 3.0SE-03 µCi/cc II) Reactor Building Exhaust PRM-RE-12 ---- --- 2.82E+1 µCi/cc --

I 0

Q)

II)

PRM-RE-13 7.50E+2 µCi/cc 7.50E+1 µCi/cc ---- ----

ra C) Turbine Building Exhaust TEA-RIS-13 8.35E-02 µCi/cc 8.35E-03 µCi/cc 8.35E-04 µCi/cc 4.22E-05 µCi/cc Radwaste Building WEA-RIS-14 3.45E-01 µCi/cc 3.45E-02 µCi/cc 3.45E-03 µCi/cc 3.98E-04 µCi/cc Exhaust Radwaste Effluent FDR-RIS-606 ---- ---- ---- 2 X HI-HI alarm

~ TSW Effluent TSW-RIS-5 ---- ---- --- 3.00E-05 µCi/cc

I C"

i Service Water Process A SW-RIS-604 ---- ---- ---- 1.00E+02 cps Service Water Process B SW-RIS-605 ---- ---- ---- 1.00E+02 cps

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 160 of 190 Mode Applicability:

1 1 1 2 1 3 4 5 1 def 1 Basis:

Threshold #1 The pre-calculated effluent monitor values presented in Table 3 should only be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available.

This EAL address gaseous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to SITE BOUNDARY doses that exceed either (ref. 1, 2):

• 100 mRem TEDE

• 500 mRem COE Thyroid The column "SAE" gaseous effluent release values in Table 3 correspond to calculated doses of 10%

of the EPA Protective Action Guidelines (TEDE or COE Thyroid) (ref. 1).

Threshold #2

• Dose assessments are performed by computer-based methods (ref. 3).

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

Escalation of the emergency classification level would be via IC RG1.

CGS Basis Reference(s):

1. Calculation NE-02-09-12 Revision 3

2. FSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling

3. PPM 13.8.1 Emergency Dose Projection System Operations

4. NEI 99-01 AS1

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 161 of 190 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 1O mrem TEDE or 50 mrem thyroid COE EAL:

RA1.1 Alert (1) Reading on any Table 3 effluent radiation monitor GT column "ALERT" for GE 15 min.

OR (2) Dose assessment using actual meteorology indicates doses GT 10 mrem TEDE or GT 50 mrem thyroid COE at or beyond the SITE BOUNDARY (Notes 1, 2, 3, 4)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes Note 4: The pre-calculated effluent monitor values presented in EALs 'RA 1.1, RS1 .1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Table 3 Effluent Monitor Classification Thresholds Release Point Monitor General SAE Alert UE PRM-RE-11 ---- ---- ---- 3.05E-03 µCi/cc U) Reactor Building Exhaust PRM-RE-12 ---- --- 2.82E+1 µCi/cc ---

s 0

Q)

U)

PRM-RE-13 7.50E+2 µCi/cc 7.50E+1 µCi/cc ---- ----

I'll

(!) Turbine Building Exhaust TEA-RIS-13 8.35E-02 µCi/cc 8.35E-03 µCi/cc B.35E-04 µCi/cc 4.22E-05 µCi/cc Radwaste Building WEA-RIS-14 3.45E-01 µCi/cc 3.45E-02 µCi/cc 3.45E-03 µCi/cc 3.98E-04 µCi/cc Exhaust Radwaste Effluent FDR-RIS-606 ---- ---- ---- 2 X HI-HI alarm

E TSWEffiuent TSW-RIS-5 ---- ---- ---- 3.00E-05 µCi/cc

s C"

J Service Water Process A SW-RIS-604 ---- ---- ---- 1.00E+02 cps Service Water Process B SW-RIS-605 ---- --- ---- 1.00E+02 cps

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 162 of 190 Mode Applicability:

1 1 1 2 1 3 4 s I def 1 Basis:

Threshold #1 The pre-calculated effluent monitor values presented in Table 3 should be used for emergency classification assessments only until the results from a dose assessment using actual meteorology are available.

This EAL address ga~eous radioactivity releases, that for whatever reason, cause effluent radiation monitor readings corresponding to SITE BOUNDARY doses that exceed either (ref. 1, 2):

• 10 *mRem TEDE

• 50 mRem COE Thyroid The column "ALERT" gaseous effluent release values in Table 3 correspond to calculated doses of 1%

of the EPA Protective Action Guidelines (TEDE or COE Thyroid) (ref. 1).

Threshold #2

• Dose assessments are performed by computer-based methods (ref. 3).

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

Escalation of the emergency classification level would be via IC RS1.

CGS Basis Reference(s):

1. Calculation NE-02-09-12 Revision 3

2. FSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling

3. PPM 13.8.1 Emergency Dose Projection System Operations

4. NEI 99-01 AA 1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A IUse Category: REFERENCE Major Rev: 033

Title:

Classifying The Emergency - Technical Bases Category:

Subcategory:

R - Abnormal Rad Release / Rad Effluent 1 - Radiological Effluent Minor Rev: N/A Page: 163 of 190 Initiating Condition: Release of gaseous or liquid radioactivity greater than 2 times the ODCM limits for 60 minutes or longer EAL:

RU1.1 Unusual Event (1) Reading on any Table 3 effluent radiation monitor GT column "UE" for GE 60 min.

OR (2) Sample analysis for a gaseous or liquid release indicates a concentration or release rate

> 2 x ODCM limits for GE 60 min.

(Notes 1, 2, 3)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes Table 3 Effluent Monitor Classification Thresholds Release Point Monitor General SAE Alert UE PRM-RE-11 ---- ---- ---- 3.05E-03 µCi/cc ti) Reactor Building Exhaust PRM-RE-12 ---- ---- 2.82E+1 µCi/cc ---

I 0

Cl) ti)

PRM-RE-13 7.50E+2 µCi/cc 7.50E+1 µCi/cc --- ----

cu

(!) Turbine Building Exhaust TEA-RIS-13 8.35E-02 µCi/cc 8.35E-03 µCi/cc 8.35E-04 µCi/cc 4.22E-05 µCi/cc Radwaste Building WEA-RIS-14 3.45E-01 µCi/cc 3.45E-02 µCi/cc 3.45E-03 µCi/cc 3.98E-04 µCi/cc Exhaust Radwaste Effluent FDR-RIS-606 ---- ---- ---- 2 X HI-HI alarm

5! TSWEffluent TSW-RIS-5 ---- ---- ---- 3.00E-05 µCi/cc

I C'

J Service Water Process A SW-RIS-604 ---- ---- ---- 1.00E+02 cps Service Water Process B SW-RIS-605 ---- ---- ---- 1.00E+02 cps

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033

Title:

Classifying The Emergency - Technical Bases Mode Applicability:

Minor Rev: NIA Page: 164 of 190

• 1 1 1 2 1 3 4 s I def 1 Basis:

Per NEI 99-01, this EAL addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways and planned batch releases from releases from non-continuous release pathways. The column "UE" gaseous release values in Table 3 represent two times the appropriate ODCM release rate limits associated with the specified monitors (ref. 1, 2, 3, 4).

The Radwaste Effluent monitor (FDR-RIS-606) Hi-Hi alarm is established per a discharge permit and should be multiplied by 2 to determine the effluent threshold.

This IC addresses a potential decrease in the level of safety of the plant as indicated by a low-level radiological release that exceeds regulatory commitments for an extended period of time (e.g., an uncontrolled release). It includes any gaseous or liquid radiological release, monitored or un-monitored, including those for which a radioactivity discharge permit is normally prepared.

Nuclear power plants incorporate design features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, and to control and monitor intentional releases. The occurrence of an extended, uncontrolled radioactive release to the environment is indicative of degradation in these features and/or controls.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

Classification based on effluent monitor readings assumes that a release path to the environment is established. If the effluent flow past an effluent monitor is known to have stopped due to actions to isolate the release path, then the effluent monitor reading is no longer valid for classification purposes.

Releases should not be prorated or averaged. For example, a release exceeding 4 times release limits for 30 minutes does not meet the EAL.

Threshold #1 - This threshold addresses normally occurring continuous radioactivity releases from monitored gaseous or liquid effluent pathways as well as radioactivity releases that cause effluent radiation monitor readings to exceed 2 times the limit established by a radioactivity discharge permit.

This EAL may also be associated with planned batch releases from non-continuous release pathways (e.g., radwaste, waste gas).

Threshold #2 - This threshold addresses uncontrolled gaseous or liquid releases that are detected by sample analyses or environmental surveys, particularly on unmonitored pathways (e.g., spills of radioactive liquids into storm drains, heat exchanger leakage in river water systems, etc.).

Escalation of the emergency classification level would be via IC RA 1.

CGS Basis Reference(s):

1. CGS Offsite Dose Calculation Manual (ODCM)

2. Calculation NE-02-09-12 Revision 3

3. 16.10.1 Radioactive Liquid Waste Discharge to the River Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033

Title:

Classifying The Emergency - Technical Bases 4.

5.

FSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling NEI 99-01 AU1 Minor Rev: N/A Page: 165 of 190

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033

Title:

Classifying The Emergency - Technical Bases Category:

Subcategory:

R - Abnormal Rad Release I Rad Effluent 1 - Radiological Effluent Minor Rev: N/A Page: 166 of 190 Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 1,000 mrem TEOE or 5,000 mrem thyroid COE EAL:

RG1 .2 General Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

• Closed window dose rates GT 1,000 mR/hr expected to continue for GE 60 min.

• Analyses of field survey samples indicate thyroid COE GT 5,000 mrem for 60 min. of inhalation.

(Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit Mode Applicability:

e 1 1 Basis:

1 2 1 3 1 4 1 5 1 def 1 Plant procedures provide guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to the EPA Protective Action Guides (PAGs). ltincludes both monitored and un-monitored releases. Releases of this magnitude will require implementation of protective actions for the public.

• Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEOE dose is set at the EPA PAG of 1,000 mrem while the 5,000 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEOE and thyroid COE.

CGS Basis Reference(s):

1. PPM 13.9.1 Environmental Field Monitoring Operations

2. NEI 99-01 AG1

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 167 of 190 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous radioactivity resulting in offsite dose greater than 100 mrem TEDE or 500 mrem thyroid COE EAL:

RS1 .2 Site Area Emergency Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

• Closed window dose rates GT 100 mR/hr expected to continue for GE 60 min.

• Analyses of field survey samples indicate thyroid COE GT 500 mrem for 60 min. of inhalation.

(Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit Mode Applicability:

e 1 1 Basis:

1 2 1 3 1 4 1 s I def 1 Plant procedures provide guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

This IC addresses a release of gaseous radioactivity that results in projected or actual offsite doses greater than or equal to 10% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public.

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 10% of the EPA PAG of 1,000 mrem while the 500 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for'TEDE and thyroid COE.

Escalation of the emergency classification level would be via IC RG1.

CGS Basis Reference(s):

1. PPM 13.9.1 Environmental Field Monitoring Operations

2. NEI 99-01 AS1

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 168 of 190 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:.

RA1.2 Alert Analysis of a liquid effluent sample indicates a concentration or release rate that would result in doses GT 10 mrem TEDE or GT 50 mrem thyroid COE at or beyond the SITE BOUNDARY for 60 min. of exposure (Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceed,ed, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit ,

Mode Applicability:

1 1 1 2 1 3 1 4 1 5 1 def 1 Basis:

• For a radiological water release, the calculated effluent concentration from a field team sample is compared to the emergency action level (ref. 1, 2, 3).

This IC addresses a release ot' gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e.g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

Escalation of the emergency classification level would be via IC RS 1.

CGS Basis Reference(s):

1. FSAR Section 11.5 Process and Effluent Radiological Monitoring and Sampling

2. PPM 13.9.1 Environmental Field Monitoring Operations

3. PPM 13.9.5 Environmental Sample Collection

4. NEI 99-01 AA 1

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A .1 Use Category: REFERENCE Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 169 of 190 Category: R - Abnormal Rad Release / _Rad Effluent Subcategory: 1 - Radiological Effluent Initiating Condition: Release of gaseous or liquid radioactivity resulting in offsite dose greater than 10 mrem TEDE or 50 mrem thyroid COE EAL:

RA1.3 Alert Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

• Closed window dose rates GT 1O mR/hr expected to continue for GE 60 min.

• Analyses of field survey samples indicate thyroid COE GT 50 mrem for 60 min. of inhalation.

(Notes 1, 2)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit

• • 1 Mode Applicability:

1 Basis:

1 2 1 3 1 4 1 s I def 1 Plant procedures, provides guidance for emergency or post-accident radiological environmental monitoring (ref. 1).

This IC addresses a release of gaseous or liquid radioactivity that results in projected or actual offsite doses greater than or equal to 1% of the EPA Protective Action Guides (PAGs). It includes both monitored and un-monitored releases. Releases of this magnitude represent an actual or potential substantial degradation of the level of safety of the plant as indicated by a radiological release that significantly exceeds regulatory limits (e;g., a significant uncontrolled release).

Radiological effluent EALs are also included to provide a basis for classifying events and conditions that cannot be readily or appropriately classified on the basis of plant conditions alone. The inclusion of both plant condition and radiological effluent EALs more fully addresses the spectrum of possible accident events and conditions.

The TEDE dose is set at 1% of the EPA PAG of 1,000 mrem while the 50 mrem thyroid COE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid COE.

Escalation of the emergency classification level would be via IC RS1.

CGS Basis Reference(s):

1. PPM.13.9.1 Environmental Field Monitoring Operations

2. NEI 99-01 AA 1

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 170 of 190 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level cannot be restored to at least the top of the spent fuel racks for 60 minutes or longer EAL:

RG2.1 General Emergency Spent fuel pool level cannot be restored to at least 0.5 ft for GE 60 min. (Note 1)

Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

The spent fuel pool is designed to maintain the water level in the pool above the top of irradiated fuel and thus providing cooling for the fuel assemblies. SFP level can be determined by FPC-LIT-21A, FPC-LIT-21 B, FPC-Ll-21 or local indication. Instrument "reference zero" is the top of the spent fuel pool racks (ref. 1) .

• The spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation to the top of the spent fuel racks. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (FPC-LIT-21A and FPC-LIT-21 B) capable of identifying SFP level near top of the fuel racks (Level 3: 0.4 ft [rounded to 0.5 ft]). (ref. 1).

This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to a prolonged uncovery of spent fuel. This condition will lead to fuel damage and a radiological release to the environment.

It is recognized that this IC would likely not be met until well after another General Emergency IC was met; however, it is included to provide classification diversity.

CGS Basis Reference(s):

1. IMDS for FPC-LIT-21A/21 B

2. NEI 99-01 AG2

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 171 of 190 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Spent fuel pool level at the top of the fuel racks EAL:

RS2.1 Site Area Emergency Lowering of spent fuel pool level to 0.5 ft Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

The spent fuel pool is designed to maintain the water level in the pool above the top of irradiated fuel and thus providing cooling for the fuel assemblies. SFP level can be determined by FPC-LIT-21A, FPC-LIT-21 B, FPC-Ll-21 or local indication. Instrument "reference zero" is the top of the spent fuel pool racks (ref. 1).

The spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation to the top of the spent fuel racks. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (FPC-LIT-21A and FPC-LIT-21 B) capable of identifying SFP level near top of the fuel racks (Level 3: 0.4 ft [rounded to 0.5 ft]) (ref. 1).

This IC addresses a significant loss of spent fuel pool inventory control and makeup capability leading to IMMINENT fuel damage. This condition entails major failures of plant functions needed for protection of the public and thus warrant a Site Area Emergency declaration.

It is recognized that this IC would likely not be met until well after another Site Area Emergency IC was met; however, it is included to provide classification diversity.

Escalation of the emergency classification level would be via IC RG1 or RG2.

CGS Basis Reference(s):

1. IMDS for FPC-LIT-21A/21B

2. NEI 99-01 AS2

• Attachment 6.1, EAL Technical Bases

I Use Category:

Number: 13.1.1A REFERENCE Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 172 of 190 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.1 Alert Uncovery of irradiated fuel in the REFUELING PATHWAY Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

The spent fuel pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel bundles.

This EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPV and spent fuel pool as well as for spent fuel pool drain down events.

This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment.

• As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EU1 .1.

Escalation of the emergency would be based on either Recognition Category R or C ICs.

This EAL escalates from RU2.1 in that the loss of level, in the affected portion of the REFUEL PATHWAY, is of sufficient magnitude to have resulted in uncovery of irradiated fuel. Indications of irradiated fuel uncovery may include direct or indirect visual observation (e.g., reports from personnel or camera images), as well as significant changes in water and radiation levels, or other plant parameters.

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

While an area radiation monitor could detect an increase in a dose rate due to a lowering of water level in some portion of the REFUEL PATHWAY, the reading may not be a reliable indication of whether or not the fuel is actually uncovered. To the degree possible, readings should be considered in combination with other available indications of inventory loss.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refuel modes.

Escalation of the emergency classification level would be via IC RS1.

CGS Basis Reference(s):

1. ABN-FPC-LOSS Loss of Fuel Pool Cooling

2. NEI 99-01 AA2

• Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 173 of 190 Category: R - Abnormal Rad Release / Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Unplanned loss of water level above irradiated fuel EAL:

RU2.1 Unusual Event UNPLANNED water level drop in the REFUELING PATHWAY as indicated by EITHER of the following:

• SFP level LE 22.3 ft.

• SFP low level alarm AND UNPLANNED rise in area radiation levels as indicated by any of the following radiation monitors:

• ARM-RIS-1 Reactor Building Fuel Pool Area

• ARM-RIS-2 Reactor Building Fuel Pool Area

• ARM-RIS-34 Reactor Building Elevation 606 Mode Applicability:

e 1 1 1 Basis:

2 1 3 1 4 1 s I def 1 The spent fuel pool is designed to maintain the water level in the pool above the top of active fuel providing cooling for the fuel bundles. The fuel pool low level alarm is actuated by level switch FP-LS-4A when fuel pool water level drops below 605' 5-1/2". SFP level is can be determined by FPC-Ll-21, FPC-LIT-21A, FPC-LIT-218 or local indication (ref. 1, 2, 3).

ifhis EAL is applicable for conditions in which irradiated fuel is being transferred to and from the RPV and spent fuel pool as well as for spent fuel pool drain down events.

ARM-RIS-1 and ARM-RIS-2 are located in the fuel pool area of the 606' elevation of the Reactor Building. ARM-RIS-34 is located on the east side of the 606' elevation of the Reactor Building (ref. 4).

This IC addresses a decrease in water level above irradiated fuel sufficient to cause elevated radiation levels. This condition could be a precursor to a more serious event and is also indicative of a minor loss in the ability to control radiation levels within the plant. It is therefore a potential degradation in the level of safety of the plant.

A water level decrease will be primarily determined by indications from available level instrumentation.

Other sources of level indications may include reports from plant personnel (e.g., from a Refuel crew) or video camera observations (if available). A significant drop in the water level may also cause an increase in the radiation levels of adjacent areas that can be detected by monitors in those locations.

The effects of planned evolutions should be considered. For example, a Refuel bridge area radiation monitor reading may increase due to planned evolutions such as lifting of the reactor vessel head or movement of a fuel assembly. Note that this EAL is applicable only in cases where the elevated reading

• is due to an unplanned loss of water level.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refuel modes.

Attachment 6.1, EAL Technical Bases

Major Rev: 033 Number: 13.1.1A \ Use Category: REFERENCE Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 174 of 190 Escalation of the emergency classification level would be via IC RA2.

CGS Basis Reference(s):

1. PPM 4.626.FPC1-2.2 (4.626.FPC2-2.2) Fuel Pool Level High/Low

2. PPM 4.627.FPC2-2.2 (4.627.FPC2-2.2) Fuel Pool Level High/Low

3. ABN-FPC-LOSS Loss of Fuel Pool Cooling

4. FSAR Table 12.3-1 Area Monitors

5. NEI 99-01 AU2

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 0~3 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 175 of 190 Category: R - Abnormal Rad Release I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.2 Alert Damage to irradiated fuel resulting in a release of radioactivity AND High alarm on any of the following radiation monitors:

• ARM-RIS-1 Reactor Building Fuel Pool Area

• ARM-RIS-2 Reactor Building Fuel Pool Area

• ARM-RIS-34 Reactor Building Elevation 606

• REA-RIS-609A-D Rx Bldg Vent Mode Applicability:

1 1 1 2 I 3 I 4 I s I def I

• Basis:

ARM-RIS-1 and ARM-RIS-2 are located in the fuel pool area of the 606' elevation of the Reactor Building. ARM-RIS-34 is located on the east sid.e of the 606' elevation of the Reactor Building (Ref. 1).

The ARM alarm setpoints are controlled by procedure.

REA-RIS-609A-D are the Reactor Building Exhaust Plenum radiation monitors. This system monitors the radiation level of the reactor building ventilation system exhaust plenum prior to its discharge from the building into the elevated release duct. A high radioactivity level in the exhaust system could be due to fission gases from damaged or leaking spent fuel or an accident (ref. 2). Actuation of the High-High alarm actuates a Secondary Containment isolation and starts SGT (ref. 3).

This IC addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment.

As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

This EAL applies to irradiated fuel that is licensed for dry storage up to the point that the loaded storage cask is sealed. Once sealed, damage to a loaded cask causing loss of the CONFINEMENT BOUNDARY is classified in accordance with EU1 .1.

Escalation of the emergency would be based on either Recognition Category R or C ICs.

This EAL addresses a release of radioactive material caused by mechanical damage to irradiated fuel.

Damaging events may include the dropping, bumping or binding of an assembly, or dropping a heavy load onto an assembly. A rise in readings on radiation monitors should be considered in conjunction with in-plant reports or observations of a potential fuel damaging event (e.g., a fuel handling accident).

Escalation of the emergency classification level would be, via IC RS 1.

Attachment 6.1, EAL Technical Bases

Number: 13.1.1 A \ Use Category: REFERENCE Major Rev: 033

Title:

Classifying The Emergency - Technical Bases CGS Basis Reference(s}:

Minor Rev: N/A Page: 176 of 190

1. CGS FSAR Table 12.3-1 Area Monitors

2. FSAR Section 11.5.2.1.2 Reactor Building Exhaust Plenum Radiation Monitoring System

3. PPM 4.602.AS-1.4 Reactor Building Exh Plenum Rad Hi-Hi

4. NEI 99-01 M2

• Attachment 6.1, EAL Technical Bases

Number: 13.1.1A

\ Use Category: REFERENCE Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 177 of 190 Category: R - Abnormal Rad Release I Rad Effluent Subcategory: 2 - Irradiated Fuel Event Initiating Condition: Significant lowering of water level above, or damage to, irradiated fuel EAL:

RA2.3 Alert Lowering of spent fuel pool level to 10 ft Mode Applicability:

1 1 1 2 1 3 1 4 1 s I def 1 Basis:

The spent fuel pool is designed to maintain the water level in the pool above the top of irradiated fuel and thus providing cooling for the fuel assemblies. SFP level can be determined by FPC-LIT-21A, FPC-LIT-21 B, FPC-Ll-21 or local indication. Instrument "reference zero" is the top of the spent fuel pool racks (ref. 1).

The spent fuel pool is equipped with primary and backup guided wave radar probes to measure pool level. The range is continuous from the high pool level elevation to the top of the spent fuel racks. Post-Fukushima order EA-12-051 required the installation of reliable SFP level indication (FPC-LIT-21A and FPC-LIT-21 B) capable of identifying SFP level providing personnel shielding (Level 2: 9.8 ft [rounded to 10 ft.]) (ref. 1).

This EAL addresses events that have caused imminent or actual damage to an irradiated fuel assembly, or a significant lowering of water level within the spent fuel pool. These events present radiological safety challenges to plant personnel and are precursors to a release of radioactivity to the environment. As such, they represent an actual or potential substantial degradation of the level of safety of the plant.

A drop in water level above irradiated fuel within the reactor vessel may be classified in accordance Recognition Category C during the Cold Shutdown and Refuel modes.

Spent fuel pool water level at this value is within the lower end of the level range necessary to prevent significant dose consequences from direct gamma radiation to personnel performing operations in the vicinity of the spent fuel pool. This condition reflects a significant loss of spent fuel pool water inventory and thus it is also a precursor to a loss of the ability to adequately cool the irradiated fuel assembles stored in the pool.

Escalation of the emergency classification level would be via ICs RS1or RS2).

CGS Basis Reference(s):

1. IMDS for FPC-LIT-21A/21B

2. NEI 99-01 AA2

• Attachment 6.1, EAL Technical Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A Tit!e: Classifying The Emergency - Tech nical Bases Page: 178 of 190 Category: R-Abnormal Rad Release / Rad Effluent Subcategory: 3-Area Radi ation Levels Initiating Condition: Radiation leve Is that IMPEDE access to equipment necessary for normal plant operation s, cooldown or shutdown EAL:

RA3.1 Alert (1) Dose rates GT 15 mR/hr in Contra I Room (ARM-RIS-19) or CAS (by survey)

OR (2) An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table 9 rooms or areas (Note 5)

Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency clas sification is warranted Table 9 Safe Operation & Shutdown Areas Room /Area Mode Applicability

• RW 467' Radwaste Control. Ro om (RHR flush to RW tanks)

RW 467' Vital Island (RHR-V-9 disconnect)

RB 422' B RHR Pump Rm (lacal pump temperatures) 3 3

3 RB 454' B RHR Pump Rm (op erate RHR-V-85B) 3 Mode Applicability:

1 1 1 2 1 3 1 4 s I def 1 Basis:

Threshold #1 The CGS Control Room requires contin uous occupancy because of its importance to assure safe plant operations and control of site security functions (Central Alarm Station).

Control Room ARM (ARM-RIS-19) measures area radiation in a range of 1 to 104 mR/hr (ref. 1).

Threshold #2 The list of plant rooms or areas with entry-related mode applicability identified 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 and shutdown. 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) are not included. In addition, the list specifies the plant mode(s) during which entry would be required for each room or area (ref. 2).

Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 179 of 190 This IC addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown. As such, it represents an actual or potential substantial degradation of the level of safety of the plant. The Emergency Director should consider the cause of the increased radiation levels and determine if another IC may be applicable.

For threshold #2, an Alert declaration is warranted if entry into the affected room/area is, or may be, procedurally required during the plant operating mode in effect at the time of the elevated radiation levels. The emergency classification is not contingent upon whether entry is actually necessary at the time of the increased radiation levels. Access should be considered as impeded if extraordinary measures are necessary to facilitate entry of personnel into the affected room/area (e.g., installing temporary shielding, requiring use of non-routine protective equipment, requesting an extension in dose limits beyond normal administrative limits).

An emergency declaration is not warranted if any of the following conditions apply.

• The plant is in an operating mode different than the mode specified for the affected room/area (i.e., entry is not required during the operating mode in effect at the time of the elevated radiation levels). For example, the plant is in Mode 1 when the radiation increase occurs, and the procedures used for normal operation, cooldown and shutdown do not require entry into the affected room until Mode 4.

• The increased radiation levels are a result of a planned activity that includes compensatory measures which address the temporary inaccessibility of a room or area (e.g., radiography, spent filter or resin transfer, etc.).

• The action for which room/area entry is required is of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

• The access control measures are of a conservative or precautionary nature, and would not actually prevent or impede a required action.

Escalation of the emergency classification level would be via Recognition Category R, C or F ICs.

CGS Basis Reference(s):

1. FSAR Table 12.3-1 Area Monitors

2. Attachment 7.4 Safe Operation & Shutdown Rooms/Areas Tables 9 Bases

3. NEI 99-01 M3

• END Attachment 6.1, EAL Technical Bases

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 180 of 190 TABLES AND NOTES Table 1 Sumps/Pool

• Any valid Hi-Hi level alarm on R-1 through R-5 sumps

• EDR GE 25 GPM

• FDR GE 10 GPM

• Wetwell level rise

• Observation of UNISOLABLE RCS leakage Table 2 AC Power Sources Offsite

• Startup Transformer TR-S

• Backup Transformer TR-B

• Backfeed 500 KV power through Main Transformers (if already aligned in modes 4, 5, def only)

Onsite

• DG1

• DG2

• Main Generator via TR-N1/N2

• Attachment 6.2, Tables and Notes

I Use Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 181 of 190 Table 3 Effluent Monitor Classification Thresholds Release Point Monitor General SAE Alert UE PRM-RE-11 ---- ---- ---- 3.05E-03 µCi/cc II) Reactor Building Exhaust PRM-RE-12 ---- ---- 2.82E+1 µCi/cc ----

s 0

Q)

II)

PRM-RE-13 7.50E+2 µCi/cc 7.50E+1 µCi/cc ---- ----

co

(!) Turbine Building Exhaust TEA-RIS-13 8.35E-02 µCi/cc 8.35E-03 µCi/cc 8.35E-04 µCi/cc 4.22E-05 µCi/cc Radwaste Building WEA-RIS-14 3.45E-01 µCi/cc 3.45E-02 µCi/cc 3.45E-03 µCi/cc 3.98E-04 µCi/cc Exhaust Radwaste Effluent FDR-RIS-606 ---- ---- ---- 2 X HI-HI alarm

~ TSW Effluent TSW-RIS-5 ---- ---- ---- 3.00E-05 µCi/cc

s C"

Service Water Process A SW-RIS-604 ---- ---- ---- 1.00E+02 cps

J Service Water Process B SW-RIS-605 ---- ---- ---- 1.00E+02 cps

• Table 4 System Plant Public Address (PA) System Communication Methods Onsite X

ORO NRC Plant Telephone System X X Plant Radio System Operations and X Security Channels Offsite calling capability from the Control X X Room via direct telephone Long distance calling capability on the X X commercial phone system

• Attachment 6.2, Tables and Notes

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033

Title:

Classifying The Emergency - Technical Bases Table 5 Safe Shutdown Areas Minor Rev: N/A Page: 182 of 190

• Vital portions of the Rad Waste/Control Building:

467' elevation vital island 487' elevation cable spreading room Main Control Room and vertical cable chase 525' elevation HVAC area

• Reactor Building

• Vital portions of the Turbine Building DEH pressure switches RPS switches on turbine throttle valves Main steam line radiation monitors Turbine Building ventilation radiation monitors Main steam line piping up to MS-V-146 and the first stop valves

• Standby Service Water Pump Houses

• Diesel Generator Building Table 7 RCS Heat-up Duration Thresholds CONTAINMENT CLOSURE RCS Status Heat-up Duration Status Intact N/A 60 min.*

established 20 min.*

Not intact not established Omin.

• If an RCS heat removal system is in operation within this time frame and RCS temperature is being reduced, the EAL is not applicable .

• Attachment 6.2, Tables and Notes

Number: 13.1.1A I Use Category: REFERENCE Major Rev: 033

Title:

Classifying The Emergency - Technical Bases Table 8 Hazardous Events Minor Rev: N/A Page: 183 of 190

• Seismic event

• Internal or external FLOODING event

• 'High winds

• Tornado strike

• FIRE

• EXPLOSION

• Volcanic ash fallout

• Other events with similar hazard characteristics as determined by the Shift Manager Table 9 Safe Operation & Shutdown Areas

• Room/Area RW 467' Radwaste Control Room (RHR flush to RW tanks)

RW 467' Vital Island (RHR-V-9 disconnect)

Mode Applicability 3

3 RB 422' B RHR Pump Rm (local pump temperatures) 3 RB 454' B RHR Pump Rm (operate RHR-V-858) 3 Table 10 Safety System Parameters

• Reactor power

• RPV level

• RPV pressure .

• Primary containment pressure

• Wetwell level

• Wetwell temperature

• Attachment 6.2, Tables and Notes

I Use Category:

Number: 13.1.1A REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 184 of 190 Table 11 Significant Transients

• Reactor scram

• Runback GT 25% thermal reactor power

• Electrical load rejection GT 25% full electrical load

• ECCS injection

• Thermal power oscillations GT 10%

• Attachment 6.2, Tables and Notes

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 185 of 190 Table 12 Notes Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release duration has exceeded the specified time limit Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for classification purposes Note 4: The pre-calculated effluent monitor values presented in EALs RA 1.1, RS 1.1 and RG1 .1 should be used for emergency classification assessments until the results from a dose assessment using actual meteorology are available Note 5: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required Note 7: This EAL does not apply to routine traffic impediments such as fog, snow, ice, or vehicle breakdowns or accidents Note 8: A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies Note 9: If the affected SAFETY SYSTEM train was already inoperable or out of service before the hazardous event occurred, then emergency classification is not warranted.

Note 1O: If the hazardous event only resulted in VISIBLE DAMAGE, with no indications of degraded performance to at least one train of a SAFETY SYSTEM, then this emergency classification is not warranted .

• END Attachment 6.2, Tables and Notes

Number: 13.1.1A J Use Category: REFERENCE Major Rev: 033 Minor Rev: NIA

Title:

Classifying The Emergency -Technical Bases Page: 186 of 190 SAFE OPERATION & SHUTDOWN AREAS TABLE 9 BASES

Background

NEI 99-01 Revision 6 ICs AA3 and HAS prescribe declaration of an Alert based on impeded access 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 HAS 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, coo/down and shutdown. Do not include rooms or areas in which actions of a contingent or emergency nature would be performed (e.g., an action to address an off-normal or emergency condition such as emergency repairs, corrective measures or emergency operations). In addition, the list should specify the plant mode(s) during which entry would be required for each room or area.

The list should not include rooms or areas for which entry is required solely to perform actions of an administrative or record keeping nature (e.g., normal rounds or routine inspections).

Further, as specified in IC HAS:

The list need not include the Control Room if adequate engineered safety/design features are 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 air intakes at different and separate locations, inner and outer atmospheric boundaries, or the capability to acquire and maintain positive pressure within the Control Room envelope .

• Attachment 6.3, Safe Operation & Shutdown Areas Table 9 Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency -Technical Bases Page: 187 of 190 The following table lists the locations into which an operator may be dispatched in order to safely shut down the reactor and reach cold shutdown conditions in accordance with plant procedures. The reason for these in-plant actions has been evaluated and a determination made whether or not the

• actions, if not performed, would prevent achieving cold shutdown. The minimum set of in-plant actions, associated locations, and operating modes to shut down and cool down the reactor are identified as "yes". These comprise the rooms/areas to be included in EAL Table 9.

If not performed, prevents Building Elevation Room Modes Reason cooldown/shutdown?

TG 441 Booster pump 1,3,4 Condensate Booster Pump No area S/D per SOP-COND-SHUTDOWN RFT Area 1,3 RFT S/D per SOP-RFT- No SHUTDOWN IR-9 Area 1 Verify Desuperheater pressure No per SOP-MT-SHUTDOWN Mech Vacuum 3 Mech Vacuum Pmp Start per No, can break vacuum and cool Pump Room SOP-AR-SHUTDOWN down with SRVs

• Mech Vacuum Pump Room OG Preheater Room Gland Exh 3

3 3

Mech Vacuum Pmp Stop per SOP-AR-START OG System S/D per SOP-OG-SHUTDOWN OG System S/D per SOP-OG-No No No Condenser Area SHUTDOWN H2 valve station 1,3,4 H2 makeup to Mn Generator No per SOP-H2/C02-0PS 501 MT Turning 1 Place MT on Turning Gear per No Gear Area , SOP-MT-START CW Pump n/a CW Pump Area 1 CW Pmp S/D per SOP-CW- No House SHUTDOWN Towers and CW 1 Monitor water level per SOP- No Basin CW-SHUTDOWN

• Attachment 6.3, Safe Operation & Shutdown Areas Table 9 Bases

IUse Category: REFERENCE Number: 13.1.1A Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 188 of 190 If not performed, Building Elevation Room Modes Reason prevents cooldown/shutdown?

RW 467 Radwaste 1,3 Remove CFDs from No Control Room service per SOP-CFD-SHUTDOWN 3 Align RW tanks to receive Yes, RWCR operator RHR water per SOP-RHR- will need to align SDC Radwaste tanks to accept RHR SDC flush water.

Vital Island 3 Close disc for RHR-V-9 per Yes, Disconnect for SOP-RHR-SDC RHR-V-9 is normally left open during power operations.

525 Communication 4 Check Oscillograph per No Room PPM 3.2.1 TMU n/a TMU Pump 1 TMU Pmp Shutdown per No Area SOP-TMU-SHUTDOWN Switch yard n/a 500KV MODs 1 Open MODs per SOP-MT- No SHUTDOWN Rx Bldg 422 B RHR Pump 3 RHR Pump local Yes, local readings Room temperature reading per of RHR pump taken SOP-RHR-SDC prior to and during flush to ensure minimal delta-Tis established 441 Railroad Bay 1 CIA N2 Bottle Change out No, Many installed per SOP-CIA-OPS bottles, infrequent task 454 B RHR Pump 3 Cycle RHR-V-85B for flush Yes, valve must be Room per SOP-RHR-SDC cycled to perform RHR SDC line flush 501 HCU Area 1 HCU Charging per SOP- No, infrequent task CRD-HCU.

548 B RHR Valve 3 Vent RHR system post No, vent not Room flush per SOP-RHR-SDC necessary to enter SDC 572 B RHR HX 3 Vent RHR system post No, vent not Room flush per SOP-RHR-SDC necessary to enter SDC

• Attachment 6.3, Safe Operation & Shutdown Areas Table 9 Bases

I Use Category: REFERENCE Major Rev: 033 Number: 13.1.1A Minor Rev: NIA

Title:

Classifying The Emergency - Technical Bases Page: 189 of 190 Table 9 Results Table 9 Safe Operation & Shutdown Areas Room/Ar.ea Mode Applicability RW 467' Radwaste Control Room (RHR flush to RW tanks) 3 RW 467' Vital Island (RHR-V-9 disconnect) 3 RB 422' B RHR Pump Rm (local pump temperatures) 3 RB 454' B RHR Pump Rm (operate RHR-V-85B) 3 Plant Operating Procedures Reviewed

1. PPM 3.2.1 NORMAL PLANT SHUTDOWN 13. SOP-MT-SHUTDOWN

2. SOP-FWH-SHUTDOWN 14. SOP-CW-OPS

• 3 . SOP-MSR-OPS

4. SOP-CW-SHUTDOWN

5. SOP-CO ND-SHUTDOWN

6. SOP-CFO-SHUTDOWN

7. SOP-TMU-SHUTDOWN

15. SOP-CG-SHUTDOWN

16. SOP-AR-START

17. SOP-MT-START

18. OSP-RHR-M 102

19. SOP-RHR-SDC

20. SOP-RCIC-SHUTDOWN

8. SOP-AS-START

9. SOP-SS-OPS 21. SOP-SS-SHUTDOWN

10. SOP-RFT-SHUTDOWN 22. SOP-H2/C02-0PS

11. SOP-RFT-OPS 23. SOP-CIA-OPS

12. SOP-AR-SHUTDOWN 24. SOP-CRD-HCU END Attachment 6.3, Safe Operation & Shutdown Areas Table 9 Bases

Number: 13.1.1A J Use Category: REFERENCE Major Rev: 033 Minor Rev: N/A

Title:

Classifying The Emergency - Technical Bases Page: 190 of 190 EMERGENCY CLASSIFICATION CHART DISTRIBUTION NOTE: The Emergency Classification Chart is provided in a separate, controlled distribution to the following locations:

Location No. Of Copies Control Room (MCR) 2 half size Control Room Simulator 2 half size Technical Support Center (TSC) 2 half size, 1 full size Alternate TSC 2 half size Emergency Operations Facility (EOF) 2 half size, 2 full size Alternate EOF 2 half size, 1 full size Joint Information Center (JIC) 1 half size Remote Shutdown Room 1 half size Simulator Remote S/D Room 1 half size NOTE: Information Only charts should be provided to the following locations:

Benton County EOC 1 half size Franklin County EOC 1 half size Washington State EOC 1 half size Grant County EOC 1 half size Adams County EOC 1 half size Yakima County EOC 1 half size END

• Attachment 6.4, Emergency Classification Chart Distribution

Initials Date Number: 13.8.1 Major Rev: 038 1 - - - - - - - - - - - - - - - - - - ~ - - - - - - - - - - - - - t Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 1 of 32 PCN#:

PLANT PROCEDURES MANUAL N/A Effective Date:

1111111111111111 IIIIII IIII IIIIII IIII IIII 13.8.1 01/17/18

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 2 of 32 DESCRIPTION OF CHANGES Implement changes due to new stack monitoring system 5 Revised step 2.12 to read "Process Radiation Monitor (PRM) Stack Monitor is".

Deleted reference to PRM-RE-1 B and 1C 7 Revised step 4.1.3.a.3) to clarify step concerning what type of monitors are available.

Deleted "other" before eDNA as redundant 10 Revised step 4.2.4 to add direction to proceed.

11 Step 4.3.1.f: Added "select "Old" and" and "currently 5/13/17" to clarify step "f'.

Deleted step "g" as redundant.

• Clarified step "h" (new step "g") concerning Met tower selection.

14 Updated References with current Engineering Calculation identifiers 15 Revised Attachment 7.5 Title to new

• 16 17 18 Revised step 2.1.1.a to replace "UE" with "Alert".

Added "or Field Team data shows a release in progress".

Step 2.1.1.d: Added "Pool" for clarification Revised Note before step 2.1.5 to stipulate that a loss of SM-8 will cause the stack monitor to be out of service.

Revised "stack monitor to "PRM Stack Monitor".

19 Revised step 3.1.1.a to change "Greater Than or Equal to" to "GE".

Added "or Field Team data shows a release is not in progress".

20 Revised step 3.1.3.b to correct selection option and clarify this step pertains to accidents in the Spent Fuel Pool.

21 Revised Note before step 3.1.6 to stipulate that a loss of SM-8 will cause the stack monitor to be out of service.

Revised "stack monitor to "PRM Stack Monitor" 22 Revised step to reflect new PRM .stack Monitor meter range 28 Revised Attachment 7.4 with updated Rad Stat Computer Points to show new computer points associated with New Stack Monitor 29 Revised Attachment 7 .5 with new Title and replaced "RE-1 B and RE-1 C" with PRM Stack Monitor. Specified which table was to be used for Table 3 equivalences and qualified to use of these equivalencies only if URI was unavailable.

Replaced old values and units in table with new values and units from Cale NE-0-09-12.

31,32 Revised Attachment 7.7 to reference PPM 11.2.24.2 instead of ODCM manual. This is now the correct reference .

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 3 of 32 TABLE OF CONTENTS Page 1.0 PURPOSE ............................................................................................................................... 4 2.0 DEFINITIONS .......................................................................................................................... 4 3.0 RESPONSIBILITIES ................................................................................................................ 6 4.0 PROCEDURE .......................................................................................................................... 6

5.0 REFERENCES

.......... :................................................................................................... :....... 14 6.0 DOCUMENTATION ...........................................................................................................:... 14 7.0 ATTACHMENTS .................................................................................................................... 15 7.1 URI User Guidance ................................................................................................................. 16 7.2 Air Sampling Worksheet Calculation ...................................................................................... 24 7.3 Obtaining Alternate Met Data ................................................................................................. 25 7.4 Computer Points Used in the eDNA View Radiological Status Screen ................................... 28 7.5 Alternate Method for PRM Stack Monitor ............................................................................... 29 7.6 Alternate Method for TEA-RIS-13 or WEA-RIS-14 ................................................................. 30 7.7 Alternate Method for ARMS or OG-RIS-612 ........................................................................... 31

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 4 of 32 1.0 PURPOSE This procedure provides instructions for the use of the computerized dose projection system Unified RASCAL Interface (URI) to predict offsite dose rates, integrated doses and radioactive material deposition for locations within the 10-mile Plume Emergency Planning Zone (EPZ) and the SO-mile Ingestion EPZ. Actual manipulation of system display terminals is described in document MAN-URl-01 referred to as the URI User's Manual.

The dose projection system is used for estimating the whole body (TEDE) and thyroid (COE) doses of onsite and offsite persons in the event of potential or actual accidental release of radioactivity to the environment. The dose projection system used at Columbia consists of a computer software program that relies on pre-calculated, real time, site-specific relationships between effluent monitor readings (or sample results) and on-site and offsite dose rates.

The dose projection system is available in the Control Room, TSC, EOF, and Alternate EOF.

Having the dose projection system software loaded on multiple, stand-alone computers located in the various emergency centers maximizes dose projection capability. Field Team data may be used to calculate dose projections or validate previous projections.

The software program supports a rapid version of dose projection using limited pathways for a rapid evaluation of the release out to 1O miles. The software also supports a detailed dose projection based on more detailed pathways with dose projections from the site boundary out to 10*or 50 miles.

• 2.0 2.1 State and county organizations will have access to this system in the EOF or by transmission of output information to their emergency centers.

DEFINITIONS COE Thyroid - Committed Dose Equivalent to the thyroid.

2.2 Delta T - The temperature difference between two sensors located at different elevations on a meteorological tower.

2.3 EDE to TEDE Ratio (EDE/TEDE) - Ratios computed by URI that are used for determining the Emergency Worker Dose Adjustment Factor. The EDE is the external gamma dose that is normally monitored by emergency workers through their self-reading dosimeters (electronic dosimeters or pocket ion chambers); and TEDE is the total dose from both EDE and the internal dose from inhaled iodines and particulates.

2.4 Elevated Release - An effluent release point model that assumes that the release point is from a discreet true elevated (tall) stack.

2.5 Emergency Worker Dose Adjustment Factor (EWDAF) - An adjustment factor determined from the EDE/TEDE Ratio that is used by offsite emergency workers to monitor their TEDE whole body dose based on their EDE doses as measured on their self-reading dosimeters. Th~

adjustment factor is the inverse of the EDE/TEDE Ratio reported by URI (for example, an EDE/TE DE Ratio of 0.15 would mean that the EWDAF is 1 / 0.15 = 6. 7) .

I Use Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 5 of 32 2.6 Radioactive Release -A radioactive release is in progress when:

Effluent monitors indicate an increase in radiation levels from normal readings for plant operating conditions, or Field teams detect environmental radiation 1O times greater than normal background.

AND The increased levels are attributable to the emergency event.

2.7 Radioactive Release Termination -A radioactive release is terminated when the following criteria have been met:

1. The source of the release has been isolated;

2. The effluent monitors are trending downward (if available);

3. Environmental Field Team surveys indicate a decrease in radiation levels or airborne radioactivity.

2.8 RASCAL - NRC supported and distributed computer software for determining source term, atmospheric dispersion, and dose consequences.

2.9 Site Boundary (SB) - Closest distance between owner controlled area boundary and core, set at 1.2 miles .

2.10 Source Term - The quantity and radionuclide makeup of the material in the release. The source term used in URI is based on NUREG-1228.

2.11 Stability Class - Values from A to G representing ranges of Delta T which in turn represent atmospheric mixing estimations. The NRC definitions of these ranges are used to define the stability classes used in URI.

2.12 Stack Door Monitor (SOM) -A portable radiation survey meter (e.g., Teletector) or radiation monitoring instrument (e.g., AMP-100) used for monitoring radiation levels at the exterior and center of the Elevated Release Stack Access Door (R-DOOR-RS 15). This provides an alternate means for monitoring Reactor Building post-accident effluent releases when the installed Process Radiation Monitor (PRM) Stack Monitor is inoperable or unavailable. SOM readings (in mR/hr) are available from HP; or, when set up for remote monitoring, through eDNA Real-Time Client or eDNA Trend from eDNA Point IDs EP99M (mid-range SOM) and EP99H (high-range SOM) in the ENW.WRM Service.

2.13 TEDE-Total Effective Dose Equivalent (TEDE)-The sum of the Deep Dose Equivalent (DOE) and the Committed Effective Dose Equivalent (CEDE).

2.14 Unified RASCAL Interface (URI) - Computer software which replaces the NRC issued RASCAL user interface for user input of dose assessment parameters and interpretation of results that interfaces to the RASCAL meteorological and dose processor modules .

I Use Category:

Number: 13.8.1 REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 6 of 32 3.0 RESPONSIBILITIES 3.1 STA I Incident Advisor Responsible for performing dose assessments until relieved by an on-call ERO Member.

3.2 Dose Projection HP Once the EOF takes control of offsite radiological assessment the Dose Projection HP is responsible for performing dose assessments until the event is terminated.

3.3 Chemistry Effluent Manager Once the TSC is manned, the Chemistry Effluent Manager is responsible for performing dose assessments if required for the TSC.

4.0 PROCEDURE 4.1 General Instructions 4.1.1 URI is not to be used for UE classification.

4.1.2 If in a declared emergency and an offsite dose or dose rate projection is needed, or if so directed, use URI to perform offsite dose calculations.

4.1.3 Access the Plant Data Information System (POIS) through the electronic Distributed Network Architecture (eDNA) software to obtain and monitor key radiation monitor, meteorological, and Plant effluent data. *

a. At a LAN supported computer:

1) Double click on the "eDNA View" icon on the Desktop, if available. If no "eDNA View" icon is available, open eDNA through the network folder and select eDNA View.

2) Select CGS, then "POIS", and "Rad Status" to obtain the "Radiological Status" screen.

• A screen print of the "Radiological Status" screen may be used to capture the current values of radiation monitor, meteorological, and effluent data for tracking changes (setup printer to Landscape mode, and use the "Print Direct" function rather than "Print" from the "File" drop-down menu on the eDNA View "Radiological Status" screen or if the file menu is not available, right click the eDNA screen and select "Print Direct").

• Access other eDNA View screens of POIS data by selecting the appropriate "eDNA View (*.rtv)" in the Application Service Utility window, as desired.

• Graphical trends of the plant parameters displayed on the eDNA View screens can be displayed by clicking in the appropriate plant parameter display box. The time span of the graphical trend may be modified by selecting "Modify Graph Parameters" in the "Graph" pull-down menu,

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 7 of 32 selecting the "Graph" tab, and entering the desired date and time range (click to check the "Y-Axis Auto Scale" box for a clearer graph).

3) Area Radiation Monitor (ARM), effluent monitor (TEA, WEA, PRM and SOM, and meteorological data are'also available through eDNA applications (e.g., eDNA Real-Time Client or eDNA Trend) from the Start menu/All Programs/eDNA folder and selecting the appropriate ENW Service (e.g., ENW.CGS, ENW.WRM, etc.).

4.1.4 Use URI to estimate doses within 15 minutes of a start of a release or as information becomes available. Also use URI to estimate offsite dose during rapidly changing meteorological conditions or release conditions as appropriate.

4.1.5 Start URI by double-clicking on the appropriate icon on the Desktop or start from the "C" drive.

4.1.6 If necessary, refer to Attachment 7.1, URI User Guidance, which provides detailed guidance on using URI.

4.1.7 Review dose projection printouts, note any qualifying factors as appropriate and brief the RPM or REM on the dose projection.

4.1.8 In the EOF/TSC, if any data is suspect, request the Radiation Detection System

• 4.1.9 Engineer or the EOF/TSC Information Coordinator to verify the data .

In the event of unmonitored release paths or if instrumentation (including alternate instrumentation) is out of service or off-scale, Field Team results are used to calculate dose projections. Use one of the following processes to assess Field T earn results: *

a. Air sample Excel spreadsheet calculator found on the Window Desktop

1) Enter the cartridge and background readings, and press the tab key to perform the calculation.

b. Attachment 7.2, Air Sampling Worksheet Calculation

1) Enter sample and background count rate and sample volume into calculations to determine micro curies/cc.

c. Air Sample Calculator in URI Detailed assessment (preferred)

1) See Attachment 7.1 for details on Air Sample Calculator use.

NOTE: GPS coordinates for the center of the Reactor Pressure Vessel are 119.33278 longitude and 46.47167 latitude

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 8 of 32 4.1.1 O Provide the Emergency Worker Dose Adjustment Factor to the REM and Field Team Coordinator for their use in establishing field team exposure limits.

a. If the Dose Adjustment Factor is 5 or greater, use Dose Adjustment Factor of 5.

b. On page three of the detailed Dose Assessment Report invert the EDE/TEDE ratios to acquire the Dose Adjustment Factor. Use the EDE/TEDE Ratio with Iodine. If respiratory protection or a thyroid blocking agent (i.e., potassium iodide) is being used, then it may be appropriate to use the EDE/TEDE Ratio without Iodine.

Example:

EDE/TEDE ratio with iodine at 2 miles is 0.90 the Emergency Worker Dose Adjustment Factor would be 1.1. (1 divided by 0.9)

c. If the EDE/TEDE ratio is less than or equal to 0.2, use a Dose Adjustment Factor of 5.

4.1.11 If intentionally venting the primary containment, perform dose projection assessment using Reactor Building Exhaust radiation monitors or field team data as applicable.

Venting is a puff release and thus release duration for intentional venting is one hour.

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: NIA

Title:

Emergency Dose Projection System Operations Page: 9 of 32 4.2 Dose Estimation Using Rapid URI NOTE: Use of Rapid URI requires that the release pathway matches one of the limited options in the Rapid URI software.

4.2.1 The following steps will provide guidance on performing a Rapid URI dose assessment with normal met tower and radiation monitors available. See Attachment 7.1 for URI implementation when normal plant indications are not available.

a. Start the URI software with icon (software on "C" drive if icon missing)

b. . If not a real event then select "This is a drill" C. Select Rapid - LIGHTNING BOLT icon at top left of startup page

d. Select Fuel Clad Damage "Yes", if EAL chart PPM 13.1.1 Table 3 Effluent Monitor reading is Greater Than or Equal to Alert value, otherwise select NO.

e. If reactor power is LE 1% (shutdown), select box for Reactor Shutdown

f. Verify shutdown date I time - change if needed

g. Select CGS 33ft Tower (Channel A primary, Channel B backup) and enter Met data, otherwise use FFTF #9

• h.

i.

j.

k.

Enter Release Duration - for ongoing release round up to next hour and add two Select Release Point Pathway - NOTE ensure correct release point pathway is selected; might have to go to "Detailed" URI Select an Effluent Monitor that has valid and on-scale data for selected pathway Enter the monit<?r reading for the selected monitor I. Enter the Release Point Flow Rate, if SGT is running and Reactor Building Exhaust Fans are shutdown; use the total of the running SGT train flowrates.

m. Select "Process Assessment" button. A green progress bar will be displayed.

Typical calculation times will be less than 20 seconds.

n. When complete, a graphic of affected areas that may require protective action will be displayed

0. Double click on the protective action display to show a pop-up of the dose results table .

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 1O of 32 NOTE: To perform an additional calculation, change any or all input values as needed and select the Process Assessment button again. If only changing one field, it may be necessary to click out of that field to get the Process Assessment button to appear.

4.2.2 See Attachment 7.1, URI User Guidance, for URI use when normal plant indications are not available.

4.2.3 Select Print to Default Printer (printer icon, upper left corner of page) to produce a paper output.

4.2.4 URI has the capability of summing dose assessments from multiple release points, proceed as follows.

a. Clicking on the sigma icon in the upper left of the screen will bring up the Assessment Summations table. Minimize this screen for later use.

b. Clicking on the plus icon in the upper right of the screen will add the latest assessment to the summations table.

c. Run a new dose assessment and click on the plus icon to sum the two dose projections. Assessments performed earlier can be added to the summations table by clicking on the browse button in summations table .

d. A combined dose assessment report and evacuation area map appears at the bottom of the summations table.

4.2.5 Have the REM or RPM compare URI output at 1.2 miles for EALs per PPM 13.1.1A and for potential protective action recommendations beyond 10 miles per PPM 13.2.2.

4.2.6 Have ED, RPM, or REM sign printed data for distribution.

a. Forward to the Emergency Director for approval prior to releasing data for distribution. 1

b. In the Control Room the Shift Manager as Emergency Director has approval authority.

c. The Washington Senior State Official approves release data for distribution during the ingestion phase.

4.2.7 Distribution of Maps and Data

a. Any dose projection maps or data printouts selected for distribution to offsite agencies shall have REM and Emergency Director review and approval.

I Use Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 11 of 32 4.3 Dose Estimation Using Detailed URI 4.3.1 The following steps will provide guidance on performing a Detailed URI dose assessment with normal met tower and radiation monitors available. See Attachment 7.1 for URI implementation when normal plant indications are not available. If in Rapid URI, you must close Rapid URI to get to Detailed.

a. Start URI with icon (software on "C" drive if icon missing).

b. If not a real event then select "This is a drill".

c. Select Detailed Assessment - CLOUD icon top left of startup page.

d. Select fuel condition - Normal Coolant or Reactor Core Accident or Spent Fuel Accident. If Normal Coolant is selected, select Spiking Factor if Reactor Bldg process radiation monitors spike following a plant transient.

e. If Reactor Core Accident - select core condition; CLAD (PPM 13.1.1 Table 3 Effluent Monitor reading is Greater Than or Equal to Alert value) or MELT (SAGs entered).

f. If Spent Fuel Accident, select "Old" and specify age based on last refueling outage shutdown, currently 5/13/17.

g. Select CGS 33ft Tower (Channel A primary, Channel B backup), if not available, then select FFTF #9 .

h. Enter Met data.

i. If reactor power is LE 1% (shutdown), select box for Reactor Shutdown.

j. Verify shutdown date I time - change if needed.

k. Enter Release Duration - for ongoing release round up to next hour and add two.

I. Double click yellow Pathway bar to show pathways selection screen.

m. Select best matching pathway.

n. Review the process reduction factor selections on the bottom of the pathway page. Defaults are specified but these may be changed as needed to better represent the plant status.

o. If pathway is through SGT, change Rx Bldg HUT to 2-24 Hours.

Factor groupings that are grey do not apply to the selected pathway.

p. Select "Accept" on the pathways page.

q. Select Monitored Release TAB center top.

NOTE: If TAB cannot be selected, then the pathway selected does not support monitored release assessment. Select a monitor that has valid and on-scale data for selected pathway.

r. Enter the Monitor reading for the selected monitor.

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 12 of 32

s. Enter the Release Point Flow Rate, if SGT is running and Reactor Building Exhaust Fans are shutdown; use the total of the running SGT train flowrates.

t. Select the 1O miles or 50 miles Process Assessment button. - A green progress bar will be displayed. Typical calculation times will be less than 20 seconds.

u. Results for dose assessment by distance will be displayed.

v. Select "View the Evacuation Area Graphic" (GLOBE) icon below "Monitored" TAB for PAR graphic.

NOTE: To perform an additional calculation, change any or all input values as needed and select the Process Assessment button again. If only changing one field, it may be necessary to click out of that field to get the Process Assessment button to appear.

4.3.2 Select Print Preview Options (magnifying glass icon directly below the "Monitored" TAB) to view the Dose Assessment report. The Receptor Point Report is not used at CGS.

4.3.3 Select Print Options (printer icon directly below the "Monitored" TAB) to print out a copy of the Dose Assessment report.

4.3.4 URI has the capability of summing dose assessments from multiple release points.

a. Selecting the sigma icon in the upper right of the screen will bring up the

• b.

c.

Assessment Summations table. Minimize this screen for later use .

Selecting the plus icon in the upper right of the screen will add the latest assessment to the summations table.

Run a new dose assessment and click on the plus icon to sum the two dose projections. Assessments performed earlier can be added to the summations table by clicking on the browse button in summations table.

d. A combined dose assessment report and evacuation area map appears at the bottom of the summations table.

4.3.5 Have the REM or RPM compare URI output at 1.2 miles for EALs per PPM 13.1.1 A and for potential protective action recommendations beyond 10 miles per PPM 13.2.2.

4.3.6 Have RPM or REM sign printed data for distribution.

a. Forward to the Emergency Director for approval prior to releasing data for distribution.

b. In the Control Room, the Shift Manager as Emergency Director has approval authority.

c. The Washington Senior State Official approves release data for distribution during the ingestion phase .

IUse Category: REFERENCE Number: 13.8.1 \

Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 13 of 32 4.3. 7 Distribution of Maps and Data

a. Any dose projection maps or data printouts selected for distribution to offsite agencies shall have REM and Emergency Director review and approval.

b. Maps selected for distribution should always be accompanied by the data. This is very important because the plume projected on the map is not closed and without the data sheet, the plume may be misinterpreted.

4.3.8 For Plume map, select "View Receptor Point Locations" (World) icon in the upper left of screen under "View"

a. Recommend selecting RASCAL puff grids and Show Balloons.

b. Hover above grid point with cursor to obtain grid projected dose.

c. Print option is in lower right hand corner of map screen .

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 14 of 32

5.0 REFERENCES

5.1 Documents 5.1.1 10 CFR 50 .47(b) Planning Standard 5.1.2 10 CFR 50, Appendix E, Emergency Plan 5.1.3 812-03-020, Elimination of Requirements for Post-Accident Sampling System 5.1.4 NUREG-0654, FEMA REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants 5.1.5 NUREG 1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents 5.1.6 RSCS TSD-13-035 (CVI 1057-00, 131) TEA and WEA replacement 5.1.7 NE-02-09-12 CGS Emergency Action Levels (EALs) Technical Bases 5.1.8 NE-02-10-05 URI 5.1.9 FW-SOFT-COTS-RASCAL and FW-SOFT-COTS-URI

• 5.1.10 5.1.11 5.1.12 CR 244578, CMR changed EAL classification value without adequate 50.54q SQA SOD SDD-PDIS-01, Plant Data Information System Software Design Description Radiation Protection Calculation 15-04, Calculation of Iodine Air Sampling Cartridge Efficiencies 5.2 Procedures 5.2.1 PPM 13.1.1A, Classifying the Emergency 5.2.2 PPM 13.2.1, Emergency Exposure Levels/Protective Action Guides 5.2.3 PPM 13.2.2, Determining Protective Action Recommendations 5.2.4 PPM 13.14.11, EP Equipment 6.0 DOCUMENTATION All logs, forms and records completed as.the result of implementing this procedure during an actual declared event shall. be retained as permanent plant records. Transmit documents to the Permanent Plant File under DIC 2304.2.

A sub-set of documents generated during drills shall be maintained in the Emergency Preparedness Department files, as necessary, to support completion of drill/exercise commitments .

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 15 of 32 7.0 ATTACHMENTS 7.1 URI User Guidance 7.2 Air Sampling Worksheet Calculation 7.3 Alternate Method for Obtaining Met Data 7.4 Computer Points Used in the eDNA View Radiological Status Screen 7.5 Alternate Method for PRM Stack Monitor 7.6 Alternate Method forTEA-RIS-13 orWEA-RIS-14 7.7 Alternate Method for ARM or OG-RIS-612

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 16 of 32 URI User Guidance 1.0 OVERVIEW 1.1 For ease of explanation, the different sections of the URI screen will be described as "groupings" with the name in the upper left of the grouping as the title of that grouping.

1.2 Options used in the Source Term Grouping will determine which options will be available in succeeding grouping.

1.3 The red octagons with the exclamation point inside are warnings that an option needs to be selected or information filled in to before the calculation can be performed. A message also appears in the lower left hand corner of the page stating that all errors must be resolved to complete the calculation. If the cursor is placed over the octagon, a message is provided that will describe the range of the input variable.

1.4 Holding the cursor over a box will often result in the program providing information about the contents of the box.

1.5 Options used in the Release Point Pathway Grouping will determine which inputs to the Process Reduction Factor will become available.

1.6 Options used in the Release Point Pathway Grouping will determine which options will be available in the Assessment Methodologies Grouping .

• 1.7

1.8 Notes

Stack Door Monitor (SOM) is a direct monitored input option. FFTF delta-Tis a direct input for stability class. CGS has committed to performing ground release assessments; so only ten meter Met conditions should be selected.

During the ingestion phase, determine manual contour lines on the 10-50 mile map by selecting grid and balloon options to determine the projected 500 µR (relocation boundary),

20 µRand 0.4 µR (food control boundary).

2.0 Rapid URI Activate Rapid URI by selecting the Rapid option (upper right hand corner) on the URI main screen. The Rapid Assessment page is now available for data input.

2.1.1 Source Term Grouping (this grouping provides input on what type of damage to the fuel has occurred)

a. In the Source Term box, answer the initial question concerning Fuel Clad Damage. If damage to the core is suspected (PPM 13.1.1 A Table 3 Effluent Monitor reading is GT Alert value), or Field Team data shows a release in progress, select "yes". This causes the assessment to include isotopic mix and inventory for fuel clad damage.

b. Selecting Fuel Clad Damage "No" causes selection of normal coolant concentrations as the basis of the source term. This also activates the Conditions for Coolant Spiking question. Select "Yes" if Reactor Bldg process radiation monitors spike following a plant transient; otherwise select "No" .

• Attachment 7.1, URI User Guidance

I Use Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 17 of 32

c. Check the "Reactor Shutdown" box if appropriate and enter date of shutdown using the pull down calendar. Overwrite the indicated time as necessary.

d. Select Damaged Spent Fuel Assembly for any Spent Fuel Pool handling accidents. This will change the Reactor Shutdown date/time box to "Last Irradiated" and allow input of when the spent fuel was last irradiated. This should be 5/13/17. (CGS does not have the ability to update software.)

2.1.2 Meteorological Data Grouping (this grouping allows input of meteorological data pertinent to the release)

NOTE: If Primary Met Tower information is not available, see attachment 7.3 for obtaining alternate met tower information. {5.1.4}

NOTE: Met Data is normally taken from the eDNA View "Radiological Status" or POIS Rad Status display. Met Channel A is primary input, Channel Bis backup input. IF any parameter cannot be obtained from eDNA View or POIS THEN refer to Attachment 7.3, Obtaining Alternate Met Data, to determine appropriate Stability Class .

• NOTE: If the wind direction value is greater than 360° subtract 360 from value before entering it into URI program.

a. Select one of the four choices of met tower input to be used. Meteorological parameters from the primary met tower are normally available on the Radiological Parameters screen. If the primary met tower (default) is not selected, the program will provide a notification of this.

b. Input Wind Speed and Wind Direction data. Wind Speed range is GEO to LE 60 mph. Wind Direction range is GE O and LT 360

c. Input delta Tor select appropriate Stability Class from the pull down menu.

Delta T range is GT -10 to LE 10.

d. Use the pull down menu to select the appropriate level of precipitation.

2.1.3 Release Duration Grouping (this grouping allows input on how long the release has been in progress)

a. Provide release duration in hours and minutes. This function defaults to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. If End of Release is not known, a default value of the time of the release is rounded up to the next hour plus two hours should be used.

b. EXAMPLE: Release has lasted for 25 minutes. Round 25 minutes up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and add 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to give a release duration of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> .

Attachment 7.1, URI User Guidance

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: NIA

Title:

Emergency Dose Projection System Operations Page: 18 of 32 2.1.4 Release Point Pathway Grouping (this grouping allows input on where the release originates from and its path for leaving the plant)

NOTE: A Radwaste building release pathway is only available in Detailed URI

a. Select the appropriate release path from the available list. A description of the pathway will appear as the cursor is held over the individual pathways. A list of available monitors will also be displayed. Choose the pathway that best matches the conditions of the release in progress.

b. Selections made in the Source Term Grouping will determine which release paths are available.

c. The release path selected will determine which effluent monitors will be available.

d. The release path selected will also determine the Process Reduction Factor (PRF) applied to the calculation of the offsite dose. IF SGT is running the Rx Bldg hold up time needs to be changed to 2-24 hours.

e. The pathway and assumptions for the input to the PRF are displayed in the lower right portion of the screen .

NOTE: In the event of a loss of SM-8, the PRM Stack Monitors will be out of service, alternate indications of Reactor Building effluents can be obtained from the Stack Door Monitor. Use the readings from the Stack Door Monitor to input directly into URI. PRMs for Turbine and RW buildings are powered from SM-7. Alternate methods for TEA and WEA monitors are the sample carts.

2.1.5 Release Point Information Grouping (this grouping allows input on the condition of the monitors at the release point)

a. Select "Yes" if effluent monitors are available. Depending on the Pathway selected, Turbine or Reactor Building monitors will be made available for selection in the Monitor Grouping.

1) Select the monitor to be used

2) Provide the appropriate reading in the Reading box, ensuring the units are correct

3) Verify the Release Point Flow Rate is correct or revise default value to accurately reflect plant conditions. If SGT is running and Reactor Building Exhaust Fans are shutdown; use the total of the running SGT train flowrates. If building (RB, RW, TB) exhaust flow indication is lost, the Control Room should be contacted to determine flow rate via ODCM 6.1.2.D method.

4) Click on the Process Assessment button to run a Rapid Assessment based on the supplied input data.

Attachment 7 .1, URI User Guidance

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 038 Minor Rev: NIA

Title:

Emergency Dose Projection System Operations Page: 19 of 32

b. Select "No" if effluent monitors are not available. This action will bring up several options for input.

a) If RCS leakage is suspected, select "Estimated RCS Leak Rate", fill in a known leak rate or select the "I Don't Know" option. Click on the Process Assessment button to run a Rapid Assessment based on the supplied input data.

b) If "Containment Leakage" is selected (only available if a Pathway from the Drywell has been selected), select either containment high rad monitor and provide an appropriate reading in the R/hr box or select "No HRA Available or Applicable". Click on the Process Assessment button to run a Rapid Assessment based on the supplied input data.

c. If Damaged Spent Fuel Assembly is selected in the Source Term Grouping and no effluent monitors are available, proceed as follows.

a) Select "Unmonitored Damaged Spent Fuel Assembly". Click on the Process Assessment button to run a Rapid Assessment based on the supplied input data.

3.0 Detailed URI Activate Detailed URI by selecting the Detailed option (upper right hand corner) on the URI main screen. The Detailed Assessment page is now available for data input.

3.1.1 Source Term Grouping (this grouping provides input on what type of damage to fuel has occurred)

a. In the Source Term box, select "Normal Coolant" if there is no core damage (No PPM 13.1.1A Table 3 Effluent Monitor reading is GE Alert value or Field Team data shows a release is not in progress). Select Spiking Factor if Reactor Bldg process radiation monitors spike following a plant transient. Tt,e numerical Spiking Factor defaults to "30". _,,

b. Reactor Core Accident is selected if there is actual or suspected damage to fuel assemblies. Technical support or core damage procedures should be used to estimate the extent of core damage.

1) The Clad damage option is selected if the core conditions have caused the fuel pin cladding to fail but the core temperature has not become sufficiently high to cause melting of the ceramic fuel matrix.

2) The .Melt option is selected if SAGs are entered and core has been uncovered for greater 30 minutes.

C. The Spent Fuel Accident option is selected if the incident involves damage to spent fuel in a depressurized condition.

1) Select "Old" for Fuel Age. "New" is not a valid option for CGS.

2) "Fuel Status" and "Amount of Spent Fuel Damage" will be provided by the software when the Selected Pathway Option is selected. This will also provide the estimated % damage based on the water level in the Spent Fuel Pool.

Attachment 7.1, URI User Guidance

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 20 of 32

3) Checking the "Unmonitored Spent Fuel Accident with No other method applicable" box is only done when there is no other option for determining spent fuel source term due to loss of building integrity, loss of monitors, or other major failure making source term assessment unavailable.

3.1.2 Meteorological Data Grouping (this grouping allows input of meteorological data pertinent to the release)

NOTE: If Primary Met Tower information is not available, see attachment 7.3 for obtaining alternate met tower information. {5.1.4}

a. Select one of the four choices of "site" met tower input to be used. The two FFTF towers are considered "site" met towers. Meteorological parameters from the primary met tower are normally available on the Radiological Parameters screen. If the primary met tower (default) is not selected, the program will provide a notification of this.

b. Input Wind Speed and Wind Direction data. Wind Speed range is GEO to LE 60 mph. Wind Direction range is GE O and LT 360. Double clicking on the Wind Direction box will bring up a depiction of a compass that may assist in selecting wind direction.

c. Input delta Tor select appropriate Stability Class from the pull down menu .

Delta T range is GT -10 to LE 10. Attachment 7.3 has additional information in determining Stability Class.

d. Use the pull down menu to select the appropriate level of precipitation.

3.1.3 Reactor Status Grouping (this grouping allows input of reactor status or time since the last time the spent fuel was irradiated)

NOTE: For ATWS Conditions:

IF reactor power is GT 1%: leave the Time Since Reactor Shutdown value set to zero unless the Main Control Room (MCR) states the reactor is shutdown.

IF reactor power is LE 1%: Contact the MCR, use the amount of time from when the MCR declares reactor shutdown.

a. When "Normal Coolant" or "Reactor Core Accident" is selected in the Source Term Grouping, "Reactor Status" will be shown next to the check box. If the Reactor is shutdown, click in this box. Change the date/time box to reflect actual shutdown if needed. The "TAS" box will update automatically. Time after shutdown can be entered directly into the "TAS" box by double clicking in the box.

b. Select Spent Fuel Accident for any spent fuel handling accidents in the Spent Fuel Pool. This will change the Reactor Shutdown date/time box to "Last Irradiated" and allow input of when the spent fuel was last irradiated. This should be 5/13/17. (CGS does not have the ability to update software.)

Attachment 7.1, URI User Guidance

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 21 of 32 3.1.4 Release Duration Grouping (this grouping allows input of the duration of the release)

I

a. Provide release duration in hours and minutes. This function defaults to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. If End of Release is not known, a default value of the time of the release is rounded up to the next hour plus two hours should be used. Release duration must be input in 15 minute increments. Range is 15 min to LE 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

EXAMPLE: Release has lasted for2'5 minutes. Round 25 minutes up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and add 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to give a release duration of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

3.1.5 Selected Pathway Options Grouping (allows input as to the release flow path and inputs to the Process Reduction Factor)

a. To view the list of available pathways, click on the small yellow box to the right of the "Pathway" box. This action will take you to the Pathways screen.

Available pathway options will depend on selections made in the Source Term Grouping. A description of the pathway will appear as the cursor is held over the "Path" box to the right of the pathway to be selected. A list of available assessment methodologies will also be displayed. Choose the pathway that matches the conditions of the release in progress.

b. The pathway chosen will determine which "Process Reduction Factors" (PRF) will be used in the dose calculation. Hol,d-up times can be changed from the default values. Other options include use of Drywell sprays, condition of the Suppression Pool, and whether SGBT is working (operable with no high moisture alarms). If SGT is running, Rx Bldg hold up time needs to be changed to 2-24hrs.

c. When pathway has been selected and PRF options have been made, click on the "Accept" button. This action will return you to the main Detailed Assessment screen. The pathway chosen will be displayed in the "Pathway" box. The PRF will be displayed under the "Pathway" box along with inputs to the PRF.

NOTE: In the event of a loss of SM-8, the PRM Stack Monitors will be out of service, alternate indications of Reactor Building effluents can be obtained from the Stack Door Monitor. Use the readings from the Stack Door Monitor to input directly into URI. PRMs for Turbine and RW buildings are powered from SM-7. Alternate methods for TEA and WEA monitors are the sample carts.

3.1.6 Assessment Methodologies Grouping (allows input to the dose assessment calculation depending on available monitors, leakage rates, samples taken, or field team results, depending on available information)

a. Monitored TAB (available when chosen Pathway is monitored by installed or temporary plant radiation monitor)

1) Release Point: The release point is based on the Selected Pathway chosen. No input required Attachment 7.1, URI User Guidance

Number: 13.8.1 J Use Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 22 of 32

2) Monitor: Select the radiation monitor you wish to use or select the building as appropriate.

3) Release Point Flow Rate: Input real time data from plant computer system/instrument readings.

Range for RB and RW bldgs. is GE O to LE 2E6 CFM, TB bldg is GEO to LE 5E6 CFM.

IF SGT is running and Reactor Building Exhaust Fans are shutdown; use the total of the running SGT train flowrates.

If building (RB, RW, TB) exhaust flow indication is lost, the Control Room should be contacted to determine flow rate via ODCM 6.1.2.D method.

4) Monitor Reading: Input real time data from plant computer system or instrument readings. Ensure units match those in the program.

Range is GE 1E-7 to LE 1E+6 µCi/cc for the PRM Stack Monitor, GE Oto LE 1E3 uCi/cc for TB and RW bldgs.

b. Containment Leakage TAB (available when chosen Pathway originates inside containment)

1) Method a) Normal Coolant /w Spiking Factor

• 2) b)

c)

% Fuel Damage range is GEO to LE 100%

Containment Radiation Monitors range is GT O to LE 1E7 R/hr Release Mode a) Leakage range is GT O to LE 100%

b) Failure to Isolate c) Catastrophic Failure d) Calculated Containment Leak Rate range is GT O cfm

c. RCS Leakage TAB (available for all chosen Pathways except Spent Fuel Pool) is used when the release pathway does not include normal plant effluent monitors (or they are unavailable).

1) Method a) Normal Coolant /w Spiking Factor b)  % Fuel Damage range is GE 1 to LE 100%, used when the Core Damage option is selected in the Source Term Grouping. The amount of core damage should be entered as obtained from TSC Core Thermal Engineer or STA.

2) Release Mode a) Unknown Release Mode b) Calculated RCS Leak Rate range is GE 1 to LT 1ES gpm Attachment 7.1, URI User Guidance

I Use Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 23 of 32

d. Release Pt. Sample TAB (available for all chosen Pathways except those directly releasing to the environment) is used when there is a valid sample collected from the effluent release point.

1) Release Point Data, no discernable range in SCFM, enter actual release point flow rate from the time of sample. If SGT is running and Reactor Building Exhaust Fans are shutdown, use the total of the running SGT train flowrates. If building (RB, RW, TB) exhaust flow indication is lost, the Control Room should be contacted to determine flow rate via ODCM 6.1.2.D method.

2) Isotope Box range is GE O to 1E4 uCi/cc, enter the observed concentrations of each identified isotope

e. Field Team TAB (available for all chosen Pathways) is used to estimate the source term and to calculate a complete dose projection based on readings obtained from monitoring teams in the field.

1) Analysis Basis a) Downwind (Miles) range .25 to 10 miles (1.2 miles is SB) b) Exp. Rate (mR/hr) range 0.1 to 99999.9 data must be based on observations in the centerline of the plume c) 1-131 Cone. (uCi/cc) range 1E-11 to 1E 11 uCi/cc, data must be based on observations in the centerline of the plume. This field is auto filled in when using the Air Sample Calculation. To use the Air Sample Calculation:

(1) Bring up Air Sample Calculation screen by selecting it from the Calculations TAB in the upper left of the URI Detailed screen.

(2) Select Count Rate Meter from the Iodine Cartridge Instrument pulldown menu.

(3) Enter Background and Cartridge Gross count rate (4) Enter Sample Flow Rate and Sample Collection Time (5) Press Enter and then the "Transfer to Field Team Cale" button (6) The Iodine concentration is transferred to the Field Team Tab for use in the assessment.

d) Survey Date/Time, enter date and time of sample taking

2) Travel Information a) "Travel Time" is derived from the "Downwind (Miles)" input and the Wind Speed used in the Meteorological Data Grouping section. No input required.

b) "Release Time" is derived from "Travel Time" and the date/time used in the "Survey Date/Time" box. No input required.

• END Attachment 7.1, URI User Guidance

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 24 of 32 Air Sampling Worksheet Calculation

1. Cartridge Filter: AgZ Filter I NOTE: 1.89 x 108 = 0.003 (eff) x 2.83 x 104 cc/ft3 x 2.22 x10 6 dpm/µCi (Sample CPM: _ _ _ _ ) - (Background CPM: _ _ _ _ ) = Net CPM _ _ __ I I

Net CPM

- - - - - - - - - - 3= µCi I cc I Activity (1.89 x 10 ) x (sample volume ji ) ----

2. Particulate Filter

• I NOTE: 6 5.65 x 109 = 0.09 (eff) x 2.83 x 104 cc/ft 3 x 2.22 x 10 dpm/µCi.

(Sample CPM: - - - - ) - (Background CPM: - - - - ) = Net CPM - - - -

Net CPM

- - - - - - - - - - 3= µCi I cc Particulate Activity (5.65 x 10 ) x (sample volume Ji ) - - - -

END

• Attachment 7.2, Air Sampling Worksheet Calculation

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 25 of 32 Alternate Method for Obtaining Met Data

1. IF Met Channel A instrument data is not available through the eDNA View "Radiological Status;'

screen, then the corresponding Met Channel B data should be used. If neither Met Channel A nor Met Channel B data is available, then alternate indications should be obtained from the following sources in the order given: {5.1.4}

• Hanford Internet Site Weather Page (primary alternate):

Select the Hanford weather icon to access the FFTF (400 Area, Station #9) meteorological information via the Internet.

IF the icon is not available, THEN start Internet Explorer and enter the following address:

http://www.hanford.gov/page.cfm/hms When the icon is selected on the desktop, either a Hanford site map or the data for FFTF will be displayed. IF the Hanford site map is displayed, THEN select the 400 Area (Station # 9) to view the FFTF data. (Use Station #11 if #9 is not available and

1. 13 last if necessary. These 3 Stations have delta T, which can be used to determine Stability Class.)

Use the wind speed and direction for the 1O meter height since a ground level release is assumed. Take 60m minus 10m temperature reading and enter into FFTF delta-Tat 10 meter tower selection.

PNNL Weather Forecaster (secondary alternate) at 373-2710 Request wind speed, direction, and differential temperature for the FFTF met tower. If this information is not available from the PNNL forecaster, contact the National Weather Service.

• Telephone the National Weather Service Forecaster (tertiary alternate) at one of the following locations:

1-541-276-8234 Pendleton, Oregon (Primary) 1-206-526-6083 Seattle, Washington (Secondary)

Request the following met data for the Hanford weather station: Wind speed (in MPH),

' wind direction, and atmospheric stability. The National Weather Service does not provide a temperature differential. The NWS will describe the stability category as neutral, moderately stable, per step 2a.

• National Weather Service website: http://www.nws.noaa.gov/

• Attachment 7.3, Alternate Method for Obtaining Met Data

Number: 13.8.1 I Use Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 26 of 32

2. IF the Stability Class is not shown on the eDNA View "Radiological Status" screen or the POIS Rad Status screen, THEN determine Stability Class as follows:

a. IF the IiT can be obtained from PN H 13-P823 Board L - Met System located in the Control Room via the Information Coordinator, THEN input !iT into URI to obtain Stability Class.

Note: The following table represents CGS IiT vs Stability Class.

Main Control Room Board L Operator Aid will require updating if the below table is revised Stability Class vs. Temperature Change for Station Met Tower Temperature Change With NRC Categories Height Stability Classification Stability (°F/212 ft- 75m -10m)

Extremely unstable A !iT s -2.2

• Moderately unstable Slightly unstable Neutral B

C D

-2.2 < !iT s -2.0

-2.0 </iTS-1.7

-1.7 < !iT s -0.6 Slightly stable E -0.6 </iTS 1.7 Moderately stable F 1.7 < !iT s 4.7 Extremely stable G !iT > 4.7

• Attachment 7.3, Alternate Method for Obtaining Met Data

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 27 of 32

b. IF no ~Tis available, THEN use the following table to determine Stability Class:

Daytime Solar Radiation (For moderate cloud cover move one column to Nighttime Conditions Surface the right)

Wind Speed Thin (mph) Spring/Fall Heavy < 3/8 Heavy Summer overcast Winter Overcast cloud Overcast Clear Sky Clear Sky (>1/2 cloud Rain cover Rain cover)

<4.5 A B B D F F D to 6.7 A B C D E F D to 11.0 B C C D D E D to 13 B D D D D D D

• >13 C D D D Table developed using guidance in EPA-454/R-99-005 (2000).

C D D END

• Attachment 7.3, Alternate Method for Obtaining Met Data

• Number: 13.8.1

• IUse Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 28 of 32 Computer Points Used In The eDNA View Radiological Status Screen RadiologiqaJ St~tus ENW.Ct;S;TIMET Reactor Power .Q012.%

Met. Channel A Met, Cltannel B DeltaT META69 degF METB69 deg F Stability Cla!ils* fylETA70 0224A MEiB70 Q224B WVindJ/ele>city META5S mph METB58 mph Wind.Direction From METAGO deg A.z METB60 degAz RxExh X213cfm ARM19 Main .. CR 0092. 111Rlhr TG Exh X198 cfm Stack; Comp x.i1s uCi/cc WV 1 606futl Pool Q093 mfl!hr COMP X499 uCi/cc Stack Lo X396uCi/cc E*34'606 .D/SEP 0085 Rlhr NG'Low. L107 uCi/cc Stacklrit X397UCi/cc N.B572SGTs** oosemRihr. Effluent X394uCits **

Stack Hi.. X:4oouCi/cc S*6 572 R.OA Q096fflRlht PlnmA )(079 mRJhr E4572 CRD Oli94mRlhr RVV Exh x3sscfm Plnmc X094 mRlhr W55.72CRD 0095 mRlhr COMP X40auCi/cc CntmfE X432Rlhr NVV"33' 501 Entry Q084Riht . NG Low L109 uCi/cc Cntmt F. X399R/hr NE7 501 TIP . 0097mRlhr Effluent. X393 uCi/s DWPress 0022:Psig NE*32 471 Elev#! 0083 Rlhr' DWTemp Q047de51 F* W24471Valve Gi099mRlhr .CntmtNG . . X469 uCi/cc FDR Sump 0041 gpm W9422RHR-A QOB7 rriRlhr

• 1 Part X470uCi/cc EDR Su"!'p 0039 51prri vir 10 422 RHR.B cioasmRlhr 2NG X398 uCi/cc N 11 422 RHR.C .aoas mRlhr 2 Part X413UCi/cc SGTSA1 X466 cfrri S23422*CRD 0098 mRlhr SGTSA2 xasscrm N '12 422 RCIC Q090mRlhr Treatment SGTSB1 X452 cfrri* E 13 422HPCS .Q091 mRlhr Pre-Treat XP7~mR/hr SGTSB2 X371 cfrri Post-Treat xossuCi/cc END Attachment 7.4, Computer Points Used in the eDNA View Radiological Status Screen

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 29 of 32 Alternate Method for PRM Stack Monitor A portable radiation survey meter or an installed radiation monitoring instrument measuring radiation levels at the exterior center of the Elevated Release Stack Access Door (R-D00R-R515) provides an alternate means of monitoring radioactive effluent releases through the Reactor Building Stack when the installed post-accident effluent monitor PRM Stack Monitor is unavailable. The radiation level readings (in mR/hr) from this Stack Door Monitor (SOM) can be input into the Monitor Reading field in URI to perform a dose projection. The technical basis for this SOM to PRM equivalency can be found in Engineering Calculation NE-02-09-12. The use of the table may be used for Table 3 equivalency if needed i.e., URI is not available. Note: for the UE value, background at 100% power is -4.1 mr/hr.

Data from eDNA Points EP99M (mid-range SOM, 5 mR/hr to 1E6 mR/hr) and EP99H (high-range SOM, 1000 mR/hr to 1E7 mR/hr) is updated approximately every minute and is displayed in mR/hr.

Always round to the nearest whole number (e.g., 23.645 mR/hr should be rounded to 24 mR/hr).

When Stack Table 3 Door Monitor Equivalent net reading is (mR/hr) (µCi/cc) Action Notes (based on filtered releases) 8 3.05E-3 (11) PPM 13.1.1A Table 3 Unusual Event EAL threshold*

12,100 2.82E+01 (12) PPM 13.1.1A Table 3 Alert EAL threshold*

PPM 13.1.1A Table 3 Site Area Emergency EAL 1.21E5 7.50E+01 (13) threshold*

1.21E6 7.50E+02 (13) PPM 13.1.1 A Table 3 General Emergency EAL threshold*

• If URI is not available.

END

• Attachment 7.5, Alternate Method for PRM Stack Monitor

Number: 13.8.1 \ Use Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 30 of 32 Alternate Method forTEA-RIS-13 orWEA-RIS-14 Compensatory Noble Gas Monitor (CNGM) sample cart(s) will be available to measure radiation levels as a compensatory measure when the normal sample racks for TEA-RIS-13 and/or WEA-RIS-14 are unavailable. There is one cart for TEA, and a different cart for WEA readings. While the equations below apply directly to the readings from either cart, it is critical to ensure that the WEA cart readings are only used for WEA dose assessments and TEA readings be used for TEA dose assessments. IF both WEA and TEA are non-functional, then both carts will be in operation and the dose assessments must be derived from the respective cart.

The technical basis for this compensatory measure equivalency can be found in RSCS TSD-13-035 (CVI 1057-00,131). The conversion equations below are used for calculating equivalent inputs for URI. Data from the CNGM will be available to the Main Control Room for monitoring a release during this compensatory measure. There is a slide at the frisker which is inserted for normal readings LTE 400,000 cpm and withdrawn for readings GT 400,000 cpm.

1) !E URI is unavailable, THEN use Table 3 of PPM 13.1.1A to classify.

2) IF frisker reading is GT 400,000 cpm, THEN direct Chemistry to withdraw the CNGM slide.

3) If calculating TEA values to input into URI, then use the following conversion factors, otherwise N/A this step:

• !E TEA frisker reading is LTE 400,000 cpm, slide inserted (Low Range Input) THEN TEA URI INPUT (µCi/cc)= [Frisker output (cpm)] / 1.01E7 cpm/µCi/cc Example: the frisker reading obtained from the sample cart is 80,000 cpm. The equivalent URI input for either TEA or WEA would be:

TEA 80000 cpm divided by 1.01 E7 cpm/µCi/cc = 7.92e-3 µCi/cc Enter this value into URI for the TEA monitor.

• !E frisker slide is withdrawn, THEN TEA URI INPUT (µCi/cc)= [Frisker output (cpm)] / 3.87E6 cpm/µCi/cc

4) If calculating WEA values to input into URI, then use the following conversion factors, otherwise N/A this step:

• !E WEA frisker reading is LTE 400,000 cpm, slide inserted (Low Range Input) THEN WEA URI INPUT (µCi/cc)= [Frisker output (cpm)] / 1.01E7 cpm/µCi/cc

• !E frisker slide is withdrawn, THEN URI INPUT (µCi/cc)= [Frisker output (cpm)] / 3.98E6 cpm/µCi/cc END Attachment 7.6, Alternate Method for TEA-RIS-13 or WEA-RIS-14

IUse Category: REFERENCE Number: 13.8.1 Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 31 of 32 Alternate Method for ARMs or OG-RIS-612 This attachment will provide direction to establish and maintain alternate method(s) for one or more inoperable ARMs or OG-RIS-612.

The following ARMs are used for EAL classification per PPM 13.1.1A: ARM-RIS-1, ARM-RIS-2, ARM-RIS-4 thru 13, ARM-RIS-19, ARM-RIS-23, ARM-RIS-24, ARM-RIS-30, ARM-RIS-32 thru 34. HP will install the alternate methods, in accordance with the requirements of PPM 11.2.9.32 and the tables within this instruction. One disabled ARM power supply can disable multiple ARMs.

OG-RIS-612 is used for EAL classification per PPM 13.1.1A. HP will establish the alternate method, as described in PPM 11.2.24.2 Attachment 8 ..5.

1.0 PORTABLE ARMS:

The wireless remote radiation monitoring devices to be used are PAM-TRX and AMP-100.

NOTE: Wireless Radiation monitors cannot be used in the Main Control Room. In the event ARM-RIS-19 is disabled, notify HP to install a non-transmitting portable area radiation monitor with an alarm set-point of 15 mr/hr.

NOTE: In the event of unplanned ARM failures, an hourly H.P. tour of the affected areas should be established until the measures of this attachment are established .

2.0 CONTROL ROOM PERSONNEL:

2.1 Direct HP to establish the applicable compensatory measures as per this attachment.

2.2 During this evolution, area radiation levels will be provided as follows:

• HP will have established wireless remote monitors into each of the areas affected, in accordance with the requirements of PPM 11.2.9.32. These monitors will provide indication to a remote monitor in the HP access area.

• In the event an area radiation level meet or exceed an ARM set-point identified by PPM 4.602.A5, the Control Room will be notified immediately. These indications should be treated as actual ARM indications. Consider directing continuous monitoring. The Control Room will be notified immediately of instrument failure.

• In the event of an emergency classification of ALERT or higher, this information will be available by request of the TSC. Continuous monitoring is not required.

• In the event that OG-RIS-612 has been replaced with an area radiation monitor, EAL classification can be made based on area radiation monitor indication that is equivalent to OG Pretreatment Hi-Hi alarm. Refer to ODCM 6.1.2.1 .

• Attachment 7.7, Alternate Method for ARMS or OG-RIS-612

Number: 13.8.1 J Use Category: REFERENCE Major Rev: 038 Minor Rev: N/A

Title:

Emergency Dose Projection System Operations Page: 32 of 32

~.o HP DEPARTMENT STAFF:

3.1 For ARMs, establish wireless remote monitors, in accordance with the requirements of PPM 11.2.9.32, in each of the affected areas, as directed by the Control Room.

3.2 For OG-RIS-612, establish the compensatory measures, as described in PPM 11.2.24.2 Attachment 8.5.

3.3 Ensure a knowledgeable individual is stationed at a monitor to observe the read-outs from each of the wireless remote monitors. Also ensure telephone communication is established with the Control Room. '

3.4 Provide direction to the knowledgeable individual to ensure the following is clearly understood:

• The radiation indications to be monitored.

• Until released by the Control Room, remain available. Check the remote monitor radiation levels at least every 15 minutes. This may be more frequent or continuous, as directed by the Control Room.

o Notify the Control Room if radiation levels on any monitor increase above an ARM set-point or 10 R/hr.

o Provide the detector locations and the radiation levels observed .

• o o

Make an HP log entry of the notification and information transmitted.

If a monitor indication fails (excluding "LOST CONTACT" for less than 10 minutes),

IMMEDIATELY notify the Control Room.

4.0 ARMs by Power Supply AR Ms 1 thru 10 powered by ARM-E/S-603A ARMs 11 thru 20 powered by ARM-E/S-6038 ARMs 21 thru 30 powered by ARM-E/S-603C ARMs 32 thru 34 powered by E-CP-H13/P614 END

• Attachment 7.7, Alternate Method for ARMS or OG-RIS-612

Initials Date Number: 13.14.11 Major Rev: 013 1 - - - - - - - - - - - - - - - - - ' - - - - - - - - - 1 Minor Rev: N/A

Title:

EP Equipment Page: 1 of 65 PCN#:

PLANT PROCEDURES MANUAL NIA Effective Date:

IIIIIII IIIII IIII IIIIII IIIII IIII IIIIII IIIII IIII IIII 13.14.11 01/17/18

I Use Category: REFERENCE Number: 13.14.11 Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 2 of 65 DESCRIPTION OF CHANGES

• • • • J4$Jlf*~~!ie~****tr~g~tf;~!t?.rm~J9*~fi~,.~I~ot*********

Revised this procedure to reflect new Rx Bldg Stack Monitor installation.

/i~~;c;j******* I( . Q~~~~I~Ji?~:1*rl:~I~~,~~i~.~mm~ry,, r~a~g,~'.i'~~Ii!atirl9 d9~~w~t!J?elical),~). {

• • • • • • r 23 Added LD-Fl-620 for instrument to indicate RCS leakage outside of Containment 27 Revised compensatory measures to reflect actual procedure HP uses to set up WRM for loss of OG-RIS-612 27 Added PRM-RE-11, 12,13. Deleted PRM-RE-1B,1C.

(New Stack Monitor has been installed).

Various Updated EALs in EAL Column to reflect EALs in PPM 13.1.1 Various Revised 10-007 Categories for numerous EPNs based on further evaluation from new EALs

I Use Category: REFERENCE Number: 13.14.11 Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 3 'of 65 TABLE OF CONTENTS Page 1.0 PURPOSE ............................................................................................................................... 4 2.0 RESPONSIBILITIES ................................................................................................................ 4 3.0 DISCUSSION ........................................................................................................................... 5 4.0 PROCEDURE .......................................................................................................................... 7 5.0 DOCUMENTATION ............................................................................................................... 10 6.0 DEFINITIONS ........................................................................................................................ 11

7.0 REFERENCES

...................................................................................................................... 14 8.0 ATTACHMENTS .................................................................................................................... 14 8.1 EP Related Important Equipment by EPN .............................................................................. 15 8.2 EP Related Important Equipment by Function ........................................................................ 32 8.3 EP Related Important Equipment - TSC ................................................................................ 42 8.4 EP Related Important Equipment - OSC ............................................................................... 46 8.5 EP Related Important Equipment - EOF (Building 34) ........................................................... 48 8.6 EP Related Important Equipment - Alternate EOF ................................................................. 57 8.7 EP Related Important Equipment- JIC .................................................................................. 60 8.8 Unplanned Loss of Equipment ............................................................................................... 64 8.9 Planned Loss of Equipment ................................................................................................... 65

I Use Category: REFERENCE Number: 13.14.11 Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 4 of 65 1.0 PURPOSE The purpose of this document is to ensure when equipment important to emergency response is removed from service for maintenance or is in a degraded condition, the correct restoration priority is assigned, compensatory measures are implemented, and the equipment is promptly restored to a functional condition.

This procedure provides identification of Important Emergency Preparedness (EP) related equipment. The equipment identified in this procedure is essential to implementation of the Emergency Plan.

For Emergency Response Facilities (ERFs) and associated equipment that are not included in the station work management process, the corrective action process is used to ensure issues are identified and corrected.

The information and direction in this procedure applies to on-site and off-site equipment and facilities normally operated or maintained by Columbia Generating Station that are necessary to meet the requirements of the Emergency Plan.

2.0 RESPONSIBILITIES 2.1 Emergency Preparedness Manager The Manager, Emergency Preparedness is responsible to assure that Important EP-related equipment is identified and repairs are executed expeditiously as necessary.

2.2 Shift Manager

(

The Shift Manager is responsible for ensuring that a Condition Report and/or Work Request is generated for Equipment identified in Attachment 8.1 and 8.2 of this procedure, and to ensure identified compensatory actions are put in place.

2.3 Work Control Manager The Work Control Manager is responsible to assure that deficient conditions of Important EP-related equipment are appropriately prioritized and scheduled to be corrected promptly in the corrective maintenance program.

2.4 Emergency Planners The Emergency Planners are responsible to assure issues with Important EP-related equipment are identified for correction .

IUse Category: REFERENCE Number: 13.14.11 Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 5 of 65 2.5 Unit Coordinator (Work Control)

Unit Coordinators are responsible to assign appropriate repair priorities to assure that Important EP equipment is factored into the daily work schedule for expeditious repair and return to service.

2.6 Corrective Action Program Manager Responsible for ensuring that the corrective action program supports the tracking and trending of deficiencies related to EP equipment.

2.7 Design Engineering Manager Responsible for ensuring that the design change process identifies any impacts to emergency plan commitments and emergency response capabilities.

2.8 Regulatory Affairs Manager Responsible for providing guidance on compliance with the station licensing basis and related reportability issues.

• NOTE: The Emergency Plan is designed as a last line of defense to address design basis accident events at a nuclear power plant, including the capability of protecting public health and safety during and following the accident. Therefore, regulations that govern EP equipment may require more timely restoration than technical specifications or other administrative controls. This procedure provides identification of Important Emergency Preparedness related equipment. The equipment identified in this procedure is essential to implementation of the Emergency Plan.

3.0 DISCUSSION 3.1 INPO 10-007, Equipment Important to Emergency Response, was used as guidance in the development of this procedure. The information contained in this procedure assists Information Technology (IT), Maintenance, and Facilities planners in identifying Important EP equipment and assigning appropriate repair priority to assure prompt return to service of any EP-related component identified here.

3.1.1 Equipment with an identified EPN is listed in Attachment 8.1.

3.1.2 IF there is no EPN, THEN reference Attachment 8.2 to determine if the affected equipment impacts the ability of the ERF to function (i.e. power, HVAC, communications).

3.1.3 Items specific to an Emergency Center are listed in Attachments 8.3 through 8.7.

These items will impact the Centers ability to function but will NOT render the center non-functional.  ;

Number: 13.14.11 I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 6 of 65 3.2 Regulatory guidance provided by the Nuclear Energy Institute (NEI) in their guidance NEI 13-01, which was endorsed by the NRC in NU REG 1022 Revision 3, differentiates between "Methods" and "Compensatory Measures," as applicable to the context in which these terms are used for equipment and facilities important to EPlan functions. Those definitions are included here verbatim as they appear in the NEI 13-01 guidance, so that their significance with respect to EPlan functions and EPlan event notifications is understood.

3.2.1 METHOD

A means that could be employed to perform an emergency response function as described in the site emergency plan or an implementing procedure described in the emergency plan. [Site emergency plans and implementing procedures typically describe primary and one or more alternate methods for performing a given function. Provided that at least one METHOD is available, then the ability to perform the associated function has not been fast.]

3.2.2 COMPENSATORY MEASURE: A temporary means, established as part of a planned activity, to perform a given emergency response function during a period when the normally used methods are unavailable such that, when implemented, there is a reasonable expectation that the function would be accomplished during an actual emergency, albeit in a possibly degraded manner. [A COMPENSATORY MEASURE need not meet the same design or operating requirements as the normally used methods but must be sufficient to support effective implementation of the site emergency plan. Also refer to the related term "VIABLE. '1

3.2.3 VIABLE

A COMPENSATORY MEASURE that (1) can restore a required function in a reasonably comparable manner and (2) is proceduralized prior to an event.

[Proceduralized means that the necessary instructions to perform a function exist in a document (e.g., a procedure, a user aid, a night or standing order, etc.) that will be followed by response personnel should an emergency occur. Further, individuals expected to implement the COMPENSA TORY MEASURE must be aware of the measure, in advance of its potential or actual implementation. A VIABLE COMPENSATORY MEASURE does not include reliance upon "skill-of-the-craft" or individual judgment.]

I Use Category: REFERENCE Number: 13.14.11 Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 7 of 65 4.0 PROCEDURE 4.1 Discovery of Degraded or Nonfunctional Equipment Important to Emergency Response (Unplanned Loss)

See Attachment 8.8, Flowchart: Unplanned Loss of Equipment Important to EP. This illustrates the process for addressing an unplanned loss of equipment, such as unplanned maintenance.

NOTE: This section pertains to unanticipated degradation or failure of equipment important to emergency response.

4.1.1 Generate a Condition Report and prioritize to restore function. This allows tracking and trending, as well as ensuring timely repair. The priority given to these actions should be commensurate with the significance of the impact on the associated emergency response function.

NOTE: Compensatory Measures identified in this procedure, which is not an Emergency Plan Implementing Procedure (EPIP), may or may not be acceptable Alternate Methods unless so recognized in another EPIP. See PPM 1.10.1, Notifications and Reportable Events, for potential notifications on loss of EP equipment.

4.1.2 Compensatory measures associated with the degraded or nonfunctional equipment can be found in the attachments to this procedure.

4.1.3 IF the equipment is INPO 10-007 Category A1 or Category A2 equipment, THEN implement identified compensatory measures.

a. The Shift Manager will implement compensatory measures to address the loss of plant equipment necessary to implement the station emergency plan. The compensatory measures should be documented in the lnop. Equip/LCO/RFO Log section of the Operations Logging System.

b. The status of out-of-service equipment can be obtained from the Operations Logging System Out of Service EPNs report.

c. Evaluate the loss of response capability. If the function requires more time to implement or cannot be implemented, evaluate the reportability requirements.

d. The Shift Manager coordinates with the licensing/regulatory affairs staff to evaluate Nuclear Regulatory Commission (NRC) reporting requirements for out-of-service EP equipment per PPM 1.10.1 .

I Use Category:

Number: 13.14.11 REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 8 of 65 4.1.4 For INPO 10-007 Category B equipment:

a. Verify the availability of redundant equipment for degraded or nonfunctional Category B equipment. If redundant equipment is not available, then the equipment should be reevaluated as Category A 1 or Category A2 equipment.

b. Category B equipment is allowed to be out of service provided redundant equipment is available to maintain the emergency response function.

c. Notify emergency response personnel of the degraded condition and the compensatory measures in effect.

4.2 Removal Of Equipment From Service For Planned Maintenance (Planned Loss)

See Attachment 8.9, Flowchart: Planned Loss of Equipment Important to EP. This illustrates the process for addressing a planned loss of equipment.

NOTE: A rigorous work management process, which includes guidance for EP equipment, is essential to ensure that maintenance is completed and that the equipment is restored to service expeditiously.

4.2.1 Individuals who are trained on and knowledgeable of emergency plan requirements should help evaluate equipment configuration changes that may impact emergency plan functions or EAL assessment capabilities.

4.2.2 For INPO 10-007 Category A1 or Category A2 equipment, ensure compensatory measures are in place before equipment is removed from service for planned maintenance. Ensure the compensatory measure is controlled in accordance with PPM 13.14.11, PPM 1.3.68, or WCl-4.

a. Evaluate the loss of response capability. If the function requires more time to implement or cannot be implemented, evaluate the reportability requirements.

b. The Shift Manager will implement the compensatory measures as soon as practical.

4.2.3 For INPO 10-007 Category B equipment, verify the availability of redundant equipment to ensure the emergency preparedness function is maintained prior to removing the equipment for planned maintenance or testing.

a. Ensure that the equipment being removed from service is not being credited as a compensatory measure for another piece of equipment.

b. When redundant INPO 10-007 Category B equipment is not available to maintain the function, the affected INPO 10-007 Category B equipment is treated as INPO 1-007 Category A equipment.

I Use Category: REFERENCE Major Rev: 013 Number: 13.14.11 Minor Rev: N/A

Title:

EP Equipment Page: 9 of 65 4.3 Degradation or Loss of Emergency Response Facilities (ERFs)

NOTE: Maintain ERFs, including alternate or backup facilities and their associated equipment, in a state of readiness. The following direction is related to planned maintenance and emergent issues and extends to all ERFs, including primary, alternate, and backup facilities as well as the equipment required to support their operation and habitability. Because of the broad scope of emergency preparedness functions conducted from these facilities, the loss of an ERF can have a significant impact on emergency plan implementation. Restoration of nonfunctional or degraded ERFs requires prompt management attention.

4.3.1 ERF Emergent Issues:

NOTE: Degraded or nonfunctioning equipment associated with these facilities is restored in a timely manner.

a. If primary power is lost to any ERF, then establish emergency power (if available).

• b. If power cannot be restored to the ERF, then contact Emergency Preparedness to establish a backup or an alternate ERF (if available).

c. If INPO 10-007 Category 8 equipment used to perform a specified ERF function is lost or degraded, then verify the availability of redundant equipment. If redundant INPO 10-007 Category B equipment is not available, then treat the equipment as INPO 10-007 Category A.

d. If INPO Category A1 or Category A2 equipment used to perform a specified ERF function is lost or degraded, then implement the appropriate compensatory measures.

e. To ensure effective implementation of the emergency plan, keep the affected members of the emergency response organization informed of changes to ERF status or availabUity and any associated compensatory measures .

IUse Category: REFERENCE Number: 13.14.11 Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 1O of 65 4.3.2 ERF Planned Maintenance

a. Before a facility or its associated support equipment is removed from service, the Emergency Preparedness manager or designee should evaluate equipment and structural configuration changes that may impact Emergency Plan functions.

b. Identify and implement compensatory measures before the facility is removed from service or is otherwise rendered uninhabitable. Minimize the out-of-service time for the facility.

c. Maintenance should inform station management, including EP, of unanticipated delays in restoration.

d. To ensure effective implementation of the emergency plan, keep the affected members of the emergency response organization informed of changes to ERF status or availability and any associated compensatory measures.

4.3.3 ERF Restoration

a. In order to apply the appropriate priority and resources to the restoration effort, the Emergency Preparedness manager or designee notifies station management when the ERF or supporting equipment cannot be restored promptly.

b. Following the restoration of ERF electrical power after an unexpected loss, the center Emergency Planner will walk down the facility to ensure the necessary support functions are available including the following:

• Habitability

• Communications

• Dose assessment

• Lighting and electrical power

• Computers and Intranet

• Technical data acquisition and display 5.0 DOCUMENTATION None

IUse Category: REFERENCE Number: 13.14.11 Major Rev: 013 1

Minor Rev: N/A

Title:

EP Equipment Page: 11 of 65 6.0 DEFINITIONS 6.1 Alternate indication - a backup means of monitoring a parameter or condition that should approximate the primary indication it is replacing.

6.2 Compensatory measure - a temporary means of mitigating the degradation or loss of an emergency response function or of maintaining the emergency response function until the equipment is restored to a fully functional condition:

6.2.1 A compensatory measure is the best available means to maintain the emergency preparedness function. Compensatory measures may include the use of redundant equipment.

6.2.2 Compensatory measures are put in place prior to scheduled equipment outages and design modifications and without delay following equipment loss or facility functional failures, to prevent or mitigate any loss of function that could result from the equipment being out of service.

6.2.3 Compensatory measures are incorporated into appropriate station processes, programs, and procedures. The station work management priority criteria appropriately address emergency preparedness equipment. Measures are in place to adjust work priorities when the compensatory measure put in place exceeds the time allowed in the evaluation or when the compensatory measure itself no longer maintains the emergency preparedness function.

6.2.4 Compensatory measures that rely on periodic monitoring also have an event-based trigger that prompts immediate and more frequent monitoring. For example, periodic sampling (such as once per shift) may be used to compensate for a nonfunctional ventilation radiation monitor. However, any increase in elevated area or airborne radiation levels in the affected buildings after the compensatory measure is put in place should trigger i~mediate and more frequent sampling.

6.3 Critical Digital Assets - Some of the components included in this document may be considered Critical Digital Assets. Contact Cyber Security Assessment Team (CSAT) for further information if necessary.

6.4 Emergency response facility (ERF) - facilities, buildings, and structures necessary to implement the site emergency plan

Number: 13.14.11 I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 12 of 65 6.5 Equipment Important To Emergency Response (EP Equipment) -

6.5.1 Systems, structures, and components, as well as essential tools and equipment, necessary to implement the emergency plan.

6.5.2 The level of detail used in determining the functionality of these items should be sufficient to allow the shift manager (SM) or station leadership team members to identify any loss or degradation of function that supports the emergency plan.

I 6.5.3 Essential tools and equipment include facility computer links to the plant computer, dedicated telephone lines, handheld radiation survey meters, air samplers, and specially equipped radiation monitoring team vehicles. The loss or degradation of these items would result in the loss of an emergency response function, as identified in the emergency plan.

6.5.4 Nonessential tools and equipment are those items that, although useful, would not result in a loss of function or diminish the emergency response capability and that are not considered equipment important to emergency response. Nonessential tools and equipment include emergency response facility fax machines, white boards, furniture, and loudspeakers.

6.6 Functional readiness -

6.6.1 The capability of emergency response facilities and EP equipment to do what they were designed to do

• 6.6.2 Consult the documents listed in the References section for regulatory guidance related to "unavailable time" and restoration timeliness.

6.7 INPO 10-007 Category A1 equipment- equipment providing the sole indication or with little or no redundancy for a parameter used to assess an emergency action level (EAL) threshold.

6.8 INPO 10-007 Category A2 equipment - equipment providing the sole means of fulfilling an emergency response function.

6.9 INPO 10-007 Category B equipment- equipment having redundant components or trains that fulfill an emergency response function or redundant indications for a parameter used to assess an EAL threshold.

6.1 O Loss of function - the inability of a facility, system, or component, including essential tools and equipment, to fulfill its emergency response purpose

Number: 13.14.11 I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 13 of 65 6.11 Maintenance actions that restore -

6.11.1 Restoring the capability of an emergency response function by repair, overhaul, or replacement.

6.11.2 This includes servicing, parts replacement, surveillance, modification, and testing as, defined in INPO AP-913, Equipment Reliability Process Description.

6.11.3 As required by INPO AP-928, Work Management Process Description, maintenance activities that impact emergency preparedness equipment, systems, or facilities must be controlled by a work management process specific to the process owner requirements.

6.11.4 This procedure applies to EP equipment that site personnel or their contractors normally control and maintain.

6.12 Plant equipment-NOTE: In some cases, the equipment required to maintain emergency preparedness systems may be outside the scope of the plant work management process. In such cases, maintenance or repair of that

• 6.12.1 equipment should be managed under processes applicable to the responsible organization.

Includes components defined by INPO AP-913, Equipment Reliability Process Description, as critical, noncritical, or run-to-failure e:12.2 Plant equipment required to maintain federal or state regulatory compliance or emergency preparedness systems will be included in this grouping. If a component does not have a criticality code assigned per AP-913, the equipment should be considered run-to-failure.

6.13 Timely restoration - Actions site personnel take to return degraded or out-of-service EP equipment to service commensurate with the significance of the associated emergency response function .

IUse Category: REFERENCE Number: 13.14.11 Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 14 of 65

7.0 REFERENCES

7.1 INPO 10-007, August 2010 Equipment Important to Emergency Response 7.2 INPO 08-007, Emergency Preparedness Manual 7.3 SWP-MAl-01, Work Management Process Overview 7.4 SWP-SEC-11, Protection of Sensitive But Un-classified Information 7.5 PPM 1.3.68, Work Management Process 7.6 01-41, Operations Work Control Expectations 7.7 IMC 0609 Appendix B, Emergency Preparedness Significance Determination Process 7.8 INPO AP-913, Equipment Reliability Process Description 7.9 INPO AP 928, Work Management Process Description 7.10 10 CFR 54(q), Code of Federal Regulations

• 7.11 NUREG 0696, Functional Criteria for Emergency Response Facilities 7.12 WCl-4, Online Work Control Processes 8.13 PPM 1.10.1 Notifications and Reportable Events 8.0 ATTACHMENTS 8.1 EP Related Important Equipment by EPN 8.2 EP Related Important Equipment by Function 8.3 EP Related Important Equipment - TSC 8.4 EP Related Important Equipment - OSC 8.5 EP Related Important Equipment -EOF (Building 34) 8.6 EP Related Important Equipment - Alternate EOF 8.7 EP Related Important Equipment - JIC 8.8 Unplanned Loss of Equipment 8.9 Planned Loss of Equipment

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: NIA

Title:

EP Equipment Page: 15 of 65 EP RELATED IMPORTANT EQUIPMENT BY EPN

.. ,. > 10~007 ... . '. ..

• EPN *< Noun Name/ EPReguirec1*Mode -- d., :S EAi.. ~lternate Indication and1 or.** .**. :*:_: :. *)))I:\? Comments

\:{:\.::

Annunciator P602.A5-3.3:

' ))t:F **

• Descri[!tiori .......... ....*.*

Off-gas Pretreatment Radiation Hi Hi Alarm.

I1 I2 I3 I I I I CategoQ!. i ,. ~:...(

.***. **>Ct>m[!ensatoQ! Measure** .******* :** ..*.,.. .......

• Establish periodic monitoring of Refer to EPN OG-RIS-612

• (*. . <

"SJAE CONDRS A1 MU4.1 OG-RR-604.

Alternate Method OUTLET RAD Refer to ODCM 6.1.2.1.

Hi Hi" Annunciator Operating Basis Establish periodic monitoring of P851.S1-5.1. Earthquake alarm. I 1 I 2 I 3 I 4 I 5 I def I H13 P823 (Board L).

"OPERATING If seismic instrumentation BASIS B HU2.1 indications are not available, use Refer to EPN SEIS-SS-1 EARTHQUAKE" information in Attachment 7.4 of ABN-EARTHQUAKE to estimate seismic intensity level.

AEA-AD-51 TSC Notify the TSC EP Planner.

M.O. Exhaust Air I 1 I 2 I 3 I 4 I 5 I def I If a FAZ signal occurs, manually Damper for AEA-FN-51 A2 NA close AEA-AD-51. Otherwise, consider relocating the TSC to the Main Control Room.

AEA-AD-52 TSC Notify the TSC EP Planner.

MO Suction Damper I 1 I 2 I 3 I 4 I 5 I def I If a FAZ signal occurs, manually for AMA-FN-52. A2 NA open AEA-AD-52. Otherwise, consider relocating the TSC to the Main Control Room.

AEA-FN-51 TSC Exhaust Fan. Notify the TSC EP Planner.

I 1 I 2 I 3 I 4 I 5 I def I A2 NA If a FAZ signal occurs, de-energize the fan at PP-TSC-1.

TSC Notify the TSC EP Planner.

AMA-AD-51 I 3 I 4 I 5 I def I I

MO Exhaust Damper I1 I2 If a FAZ signal occurs, manually for AMA-FN-52 A2 NA open AMA-AD-51. Otherwise, consider relocating the TSC to the Main Control Room.

AMA-CF-52 TSC Notify the TSC EP Planner.

AMA-FU-52 Charcoal I 1 I 2 I 3 I 4 I 5 I def I Establish periodic radiological Filter/Adsorber A2 NA monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 16 of 65

• t. 5;Nc:fU~Nafuel rriate!lndicadon***aridla*

• * * <Descriutio 1compensatoni*MeasUre/

AMA-FL-52 TSC Notify the TSC EP Planner.

AMA-FU-52 Pre filter Establish periodic radiological A2 NA monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

AMA-FN-51 AMA-AH-51 I1 12 13 14 15 I def I Notify the TSC EP Planner Recirculation Fan Establish alternate cooling as A2 NA needed. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

AMA-FN-52 TSC I1 !2 13 14 /5 i def I Notify the TSC EP Planner.

Emergency Supply Fan Establish periodic radiological A2 NA monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

AMA-FU-52 TSC I1 /2 13 14 i s I def I Notify the TSC EP Planner.

Emergency Establish periodic radiological HEPA/Carbon Filter A2 NA monitoring of the TSC. If a FAZ Unit occurs, consider relocating the TSC to the Main Control Room AMA-HF-52 TSC I1 !2 13 14 15 I def I Notify the TSC EP Planner.

AMA-FU-51 HEPA Establish periodic radiological Filter A2 NA monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

AOA-AD-51 TSC i 1 /2 13 14 i s I def I Notify the TSC EP Planner.

M.O. Outside Air If a FAZ signal occurs, manually Damper A2 NA close AMA-AD-51. Otherwise, consider relocating the TSC to the Main Control Room.

AOA-EHC-51 TSC I1 12 13 14 15 I def I Notify the TSC EP Planner.

AMA-AH-51 Outside A2 NA Establish alternate Air Duct Heater cooling/heating as needed.

Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 17 of 65

. e~ R~gujred Mode '"" 10:::()()7 ? tter11atelndication*and>ttor catego~. ..* CompensatoryMe*asure Y AOA-FN-51 TSC Remote Air Intake 1 12 13 14 15 I def I Notify the TSC EP Planner.

Fan Establish periodic radiological A2 NA monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

AOA-V-51A TSC Remote Air Intake Isolation Valve 12 13 14 is I def I Notify the TSC EP Planner.

If a FAZ occurs, manually open A2 NA AOA-V-51A or 51 B, as appropriate. Otherwise, consider relocating the TSC to the Main Control Room ..

AOA-V-518 TSC Remote Air Intake Isolation Valve 12 13 14 is I def I Notify the TSC EP Planner.

If a FAZ occurs, manually open A2 NA AOA-V-51A or 518, as appropriate. Otherwise, consider relocating the TSC to the Main Control Room.

ARE-CR-51 TSC Chiller 12 13 14 Is 1 def I Notify the TSC EP Planner.

Establish alternate cooling as A2 NA needed. Establish periodic temperature monitoring of the TSC.

ARE-TCV-1 TSC 12 13 14 15 I def I Notify the TSC EP Planner.

AMA-CC-51 Establish alternate Temperature Control A2 NA cooling/heating as needed.

Valve (Expansion Establish periodic temperature Valve) monitoring of the TSC. Verify ARE-TCV-2 is functional.

ARE-TCV-2 TSC I1 12 13 14 I5 I def I Notify the TSC EP Planner AMA-CC-52 Establish alternate Temperature Control A2 NA cooling/heating as needed.

Valve (Expansion Establish periodic temperature Valve) monitoring of the TSC. Verify ARE-TCV-1 is functional.

Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 18 of 65

/ 'NotiniN~m Y>/Description ARE-TIC-51 TSC Notify the TSC EP Planner.

Temperature Establish alternate cooling as Indicating Controller A2 NA needed. Establish periodic For ARE-CR-51 temperature monitoring of the TSC. If needed, establish alternate temperature control.

Notify the TSC EP Planner.

ARE-TIS-52 TSC AOA-AD-51 Supply I 1 I 2 I 3 I 4 I 5 I def I A2 NA Establish alternate cooling/heating as needed Plenum Temp Switch ARM-RIS-1 Reactor Building 606' I 1 I 2 I 3 I 4 I 5 I def Fuel Pool Area ARM-RIS-1 / 2 RU2.1 OR B

ARM-RIS-2 Reactor Building 606' I 1 I 2 I 3 14 1* 5 I def RA2.2 HP monitoring during load movement over spent fuel Fuel Pool Area Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 19 of 65 ARM-RIS-4 Reactor Building 522' 1 2 3 1 4 5 East CRD Area ARM-RIS-5 Reactor Building 522' 2 3 14 5 CG1.1 West CRD Area CG1.2 ARM-RIS-6 Reactor Building 572' B 1 2 3 4 5 H2 Recombiner Area FS1.1 ARM-RIS-7 Reactor Building 501' 2 3 FA1.1 4 5 Tl P Drive Area ARM-RIS-8 Reactor Building 572' 2 3 4 5 SGT System Area

_')

ARM-RIS-9. Reactor Building 422' RA3.1 Alternate Method 1 2 3 4 15 def 1 "A RHR Pump Room Refer to PPM 13.8.1 CG1.1 RCS Potential Loss ARM-RIS-10 A1 CG1.2 Reactor Building 422' "B" RHR Pump Room FS1.1 FA1.1 ARM-RIS-11 Reactor Building 422' "C" RHR Pump Room I1 I2 I3 I4 I5 I CG1.1 B CG1.2 ARM-RIS-12 Reactor Building 422' RCIC Pump Room FS1.1 ARM-RIS-13 Reactor Building 422' FA1.1 HPCS Pump Room Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 20 of 65

• • • • .*1TEP;R~goiredMoaer: ***** * * . . nv1o;;(j(j7************ *nerriatelni:ltcationaiialor.*i*****

• ** . ** * ** \Oescription ,* Compensatory:Meas1.1re*.*. **

• ARM-RIS-19 Rad Waste Building RA3.1 Non-wireless Area Radiation 501' I 1 I 2 I 3 I 4 I 5 I def A1 Monitor Main Control Room ARM-RIS-23 Reactor Building 422' CG1.1 CRD Pump Room CG1.2 Refer to PPM 13.8.1 Alternate Method ARM-RIS-24 Reactor Building 471' B FS1.1 1 2 3 4 5 5.3.1 Table 24 RCS Potential Loss Northwest Area I I I I I FA1.1 ARM-RIS-30 Radwaste Control Rm A1 RA3.1 Refer to PPM 13.8.1 ARM-RIS-32 Reactor Building 471' CG1.1 High Range ARM CG1.2 Refer to PPM 13.8.1 ARM-RIS-33 Reactor Building 501' B FS1.1 1 2 3 4 5 5.3.1 Table 24 Alternate Method High Range ARM I I I I I FA1.1 ARM-RIS-34 Reactor Building 606' RCS Potential Loss RU2.1 ARM-RIS-1/2 High Range ARM OR B RA2.2 HP monitoring during load movement over spent fuel Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 21 of 65 tefriate'lridicatioriiand/*c>I' \, , .

omt2ensatofi:Measure \

CMS-LR-3 Suppression Pool CA1.1 Level Recorder 12 I3 14 I5 I CU1.1 CG1.2 B CS1.2 Redundant Instrument HCTL MA3.1 CMS-LR-4 Suppression Pool Level Recorder 12 13 14 I5 I MU3.1 MS6.1 CMS-PR-1 Primary Containment Pressure I1 12 I3 14 I5 I CG1.1 CMS-PR-2 Primary Containment Pressure 1 12 13 14 I5 I CG1.2

• PC Loss B FA1.1 Redundant Instrument

• RCS Loss CMS-PR-3 Wetwell Pressure

• I2 13 14 I5 I FS1.1 HCTL FG1.1 CMS-PR-4 Wetwell Pressure I 2 13 14 I5 I CMS-RIS-27E In-Containment Hi Range Area Radiation I1 12 13 I I I

• PC Potential Loss FA1.1 Readout CMS-RIS-27 A/8 OR Perform RCS Loss FS1.1 "

CMS-RIS-27F In-Containment Hi 1 12 13 B Core Damage Estimate if needed

• Fuel Clad Loss Range Area Radiation I I FG1.1

• URI Readout Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 22 of 65

, . . *.*.:.,* {. ~e~~~!tf/n1*. r*.* ...... ..

'ji!Ii( . .~*i**********i;:

y

.L***EPrReijuired Mode

• .:*:+/\'I'.

<.>10-001*

Categoci /: lI:! --*--,: :: ; '.KA!terriate CAi ...i/.c I Indication and for{ .* ~*?.

• . Comlieiisafod'. Measlire ..

0

rrt~~~ ;*.*.* .*,*. *. .

CMS-SR-13 H2/02 Analyzer I I I I4 I5 I I CG1.1 B CG1.2 CMS-SR-14/13

• PC Potential Loss CMS-SR-14 H2/02 Analyzer I I I I4 I5 I I CMS-TR-5 Wetwell Temp MA3.1 I1 I2 I3 I I I I B MU3.1 Redundant Instrument

• PC Potential Loss CMS-TR-6

• HCTL MS6.1 EDR-FRS-623 Reactor Bldg CA1.1 EDR-FT-37 and I 1 I 2 I3 I4 I5 I I CU1.1 Drywell FDR-FT-38 B Manual Determination Recorder MU5.1 E-LP-TSC1 TSC Lighting Power Notify the TSC EP Planner.

Supply I 1 I 2 I 3 I 4 I 5 I def I A2 NA Relocate TSC E-MC-7AB TSC power supply. Notify the TSC EP Planner.

I 1 I 2 I 3 I 4 I 5 I def I A2 NA Relocate TSC E-PP-TSC1 TSC Power Panel Notify the TSC EP Planner.

I 1 I 2 I 3 I 4 I 5 I def I A2 NA Relocate TSC E-VM-DPS1/1 125 VDC VITAL DIST CU4.1 PNL S1/1 I1 I2 I3 I4 I5 I I Local voltmeter or on back of A1 MS2.1 remote indication meter.

MG1.2 E-VM-DPS1/2 125 VDC VITAL DIST CU4.1 '

PNL S1/2 I1 I2 I3 I4 I5 I I Local voltmeter or on back of A1 MS2.1 remote indication meter.

MG1.2 Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11 t - - - - - - - - - - - - - - - - - - - - ~ - - - - - - - - - - - - - - - ! Minor Rev: N/A

• Use Category: REFERENCE Major Rev: 013

Title:

EP Equipment Page: 23 of 65

<***.*.*****J*******~t****\:*:***t}

,:;;.=*.,.;: ;_..

. .:* *.'* <'NounNamet'*

>oescriptioil.'*

,e .*.*. ***'********.~.*...,..o......d.*.:.,.*,..*..*,e..*..*.*. **,' '., ** ,'*:*. >

• : + *.*.*.*.:. .\?.*,*.*':******.**.**.'*E******p**.*.**,***R********e*:* qu**,*.* * :i.:'*i'.:.*.:.d *10:001 jt:;ategoryi; . i. ,.,/* * *....EALi I JAlterriate* tndicatiort'and

• ,*. ~>Compensatory Measure / ;i : ':i;:; Comments I or*:.+~ I*. . . / ***.*. . . .*. :*

• 2-;,. i* U FDR-RIS-606 Radwaste Liquid Effluent Rate meter I. I 2 I 3 I 4 I 5 I def I A1 RU1 .1 Perform Sample and Verify Release Rate Calculation Prior to RA 1*2 Release.

RHR "A" pump room FDR-LS-41 level high annunciator I2 I3 I4 Is I FDR-LS-42 RHR "B" pump room 1 1 j 2 j 3 I 4 j 5 j level high annunciator B

CG1.2 FDR-LS-43 RHR "C" pump room level high annunciator CS1.2 CA6.1 Visual once per shift FDR-LS-44 RCIC pump room level high annunciator MA8.1 HU3.2 FDR-LS-45 LPCS pump room level high annunciator j 1 I. 2 I. 3 I. 4 I. 5 .

j FDR-LS-46 FPC-Ll-21 ~~:;t 1

Fuel Pool Water j 1 j 2 j 3 I4 j 5 j def j B RG2.1 RS2.1 FPC-LIT-21A, 21B or 21 OR FPC-LIT-21A ~~:; t 1

Fuel Pool Water I 2 I 3 I 4 I 5 I def I RA2.1 Visual twice per Shift OR Prefered compensatory measure is a level Skimmer Surge Tank Low Level indicator RA2.3 Not In With FPC Flow FPC-LIT-21 B ~~:;t 1

Fuel Pool Water j 1 j 2 j 3 I4 j 5 j def j RU2.1 LD-Fl-620 B MU5.1 System flow evaluation Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 24 of 65 LD-TE-3NB RWCU-P-1A Rm LD-TE-3C/D RWCU-P-1B Rm I 1 I2 I3 I LD-TE-4NB RCIC Pump Rm I1 I2 I3 I LD-TE-18NB RHR-P-2B Rm I1 J 2 13 I I I LD-TE-18C/D RHR-P-2A Rm I1 I2 I3 I LD-TE- RHRAHXRm 18E/F/G/H I1 I2 I3 I FA1.1 LD-TE- RHRBHXRm 18J/K/UM J 1 I2 13 I B FS1.1 Use Leak Detection temperature

• PC Loss element in same vicinity

• RCS Potential Loss FG1.1 LD-TE-24NB RWCU Pipe Area RB 548 N (R509) I1 I 2 13 I LD-TE-24C/D RWCU Pipe Area RB 548 S (R511) I 1 I 2 13 I I I RWCU Pipe Area RB LD-TE-24E/F 522 N (R408) I1 I2 I3 I I I LD-TE-24G/H Above RWCU Pump Rooms RB 522 (R409) I 1 I 2 13 I LD-TE-24J/K TIP Mezzanine RB 501 NE (R313) I1 I2 I3 I LD-TE- Main Steam Tunnel 31NB/C/D I1 I2 I3 I Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 25 of 65 lt$J~J:~::::~~~:~gr~t.' C'"'

Channel B Me_t Tower instruments FFTF is Alternate B N/A METHOD Hanford FFTF Website Backup Control Rack Channel A Met Tower MET-CPU-18 CPU 12 13 14 is I def I instruments FFTF is Alternate B N/A

/ METHOD Hanford FFTF Website Primary 245' 245' temperature on Channel B MET-TE-10A Temperature Element I1 12 13 14 is I def I on eDNA/PDIS FFTF is Alternate B N/A METHOD Hanford FFTF Website Backup 245' 245' temperature on Channel A MET-TE-10B Temperature Element 12 13 14 15 I def I on eDNA/PDIS FFTF is Alternate B N/A METHOD Hanford FFTF Website Primary 33' 33' temperature on Channel B on MET-TE-11A Temperature Element 12 13 14 is I def I B N/A eDNA/PDIS FFTF. is Alternate Hanford FFTF Website METHOD Backup 33' 33' temperature on Channel A on MET-TE-11B Temperature Element 12 13 14 is I def I eDNA/PDIS FFTF is Alternate B NIA METHOD Hanford FFTF Website Temperature Recorder Delta temperature (245'-33') on MET-TR-1 1 12 13 14 15 I def I Channel A or B on eDNA FFTF is Alternate B NA METHOD Hanford FFTF Website Wind Direction 33' wind direction on Channel A MET-WDR-4 Recorder 12 13 14 1 s I def I or Bon eDNA FFTF is Alternate B NA METHOD Hanford FFTF Website MET-WMON-2A Primary 33' Wind 33' wind speed/direction on Speed/Direction I1 12 13 14 is I def I Channel B on eDNA/PDIS FFTF is Alternate B N/A Monitor METHOD Hanford FFTF Website Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: NIA

Title:

EP Equipment Page: 26 of 65 r~**: . ** }:;; ;(}Noun 'Namrti'f

)  ::::*.*

• '.'+:Attetnate:1ndicatiori'and**1 or

• • rt**t;;;;<Pitf~~

EP:Reguired Mode 1()~007'\ p;*,rti=AEC //

• *.f.*\ Descri!;!tiomk ':.\**-; iCatego!i 1*.*1.i***.*******r. ;.**** . . . 1*. .**.,**.** yJCom~erisato!:£'.Measure****t !****.***

iji*******iI*** .:n!' \'

MET-WMON-2B Backup 33' Wind 33' wind speed/direction on Speed/Direction I 1 I 2 I 3 I 4 I 5 I def I N/A Channel A on eDNA/PDIS FFTF is Alternate Monitor B METHOD Hanford FFTF Website MET-WSR-4 Wind Speed Recorder CA6.1 33' wind speed on Channel A or I1 I2 I3 I4 I5 I I B Bon eDNA FFTF is Alternate MA8.1 METHOD Hanford FFTF Website MS-LR-615 RPV Level - Fuel Zone CG1.1 I1 I2 I3 I4 I5 I I CS1.1 CS1.2

• TAF CA1.1

• SAG Entry B Redundant Instrument FS1.1

• RCS Loss FG1.1

• Fuel Clad Loss MG1.1 MS6.1 MS-LR/PR-623A RPV Press & Level Recorder (Post Ace I I I I4 I5 I I CktA) CS1.1 Redundant Instrument Potential HCTL MS-LR/PR-623B Post Accident Mon B CA1.1 4 5 Instr Loop B Press & I I I I I I I Level Recorder OFMA-HF-1C HEPA Filter Bank for AHU1 Cold Deck I1 I2 I3 I4 I 5 I def I A2 NA Notify EOF Emergency Planner EOF confirms habitability upon reporting and relocates if necessary OFMA-HF-1H HEPA Filter Bank for EOF confirms habitability AHU1 Hot Deck I 1 I 2 I 3 I 4 I 5 I def I A2 NA Notify EOF Emergency Planner upon reporting and relocates if necessary OFMA-HF-3 EHPA Filter bank for EOF confirms habitability supply fan 3 I1 I2 I3 I I I I A2 NA Notify EDF Emergency Planner upon reporting and relocates if necessary Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 27 of 65 iifti*)Noun****Namet>  : l'.Alternate***tildication and For<

Description Compensatory Measure OG-RIS-612 Off-Gas Pretreatment Refer to EPN -

Rad Rate meter Annunciator P602.AS-3.3:

Direct HP to establish WRM per "SJAE CONDRS OUTLET A1 MU4.1 PPM 11.2.24.2 PPM 13.8.1 RAD Hi Hi" Alternate Method PRM-DPT-2 PRM-SQRT-2 Reactor Building Flow Rate Instruments 12 13 14 is I def I RG1.1 RS1.1 Alternate Method ODCM PRM-Fl-2 A1 Refer to ODCM 6.1.2.D RA1.1 6.1.2.D RU1.1 PRM-RE-11 Reactor Building I1 12 13 14 15 I def I Direct HP to establish Stack Door Exhaust, Low Monitor (SDM) per PPM 11.2.24.2 -- radiation monitoring RU1.1 SDM is Alternate A1 at the RB Stack Access Door (R-METHOD DOOR-RS1 S). SDM readings are obtained from HP or from eDNA Point EP99M in ENW.WRM.

PRM-RE-12 Reactor Building 12 13 14 15 I def I Direct HP to establish Stack Door Exhaust, Intermediate Monitor (SDM) per PPM 11.2.24.2 -- radiation monitoring RA1.1 SDM is Alternate A1 at the RB Stack Access Door (R-METHOD DOOR-RS1S). SDM readings are obtained from HP or from eDNA Point EP99M in ENW.WRM.

PRM-RE-13 Reactor Building 12 13 14 15 I def I Direct HP to establish Stack Door Exhaust, High Monitor (SDM) per PPM RG1.1 11.2.24.2 -- radiation monitoring RS1.1 at the RB Stack Access Door (R- SDM is Alternate A1 DO_OR-RS1 S). SDM readings are METHOD obtained from HP or from eDNA Points EP99M or EP99H in ENW.WRM.

REA-R1S-609A-D Reactor Building Vent 12 13 14 is I def I B RA2.2 Redundant Instrument Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: NIA

Title:

EP Equipment Page: 28 of 65

:?i>>Noun Nami:f/ **. Alternate lridic~tid11 and For

\., Descriptiori ncompensatoryMeasure . .

• RFW-Pl-605 Reactor Pressure B CA3.1 Redundant Instrument

• PC Potential loss (0-1200 psig)

• HCTL RWCU-Tl-607 RWCU-P-1A, 1B Discharge Header 14 I5 I B CA3.1 Temporary Thermocouples or RPVRTDs RRC-TR-650 RHR-Only accurate if an RRC pump is running and Temp CU3.1 RWCU flow is greater TRS-601 than 50 gpm.

I 5 I def I SEIS-SS-1 Triaxial Acceleration 1 12 13 14 Establish periodic monitoring of Refer to EPN-Seismic Switch HU2.1 H13 P823 (Board L red lights). Annunciator P851.S1-5.1:

A1 Refer to PPM ABN-Earthquake if "OPERATING BASIS full instrumentation is lost. EARTHQUAKE" SGT-FT-1A1 SGT A-1 Flow 3 def I1 2 4 I5 B NA Field Team Results SGT-FT-1A2 SGT A-2 Flow 2 3 def I1 4 I5 B NA Field Team Results SGT-FT-1 B1 SGT B-1 Flow 2 3 4 is def B NA Field Team Results SGT-FT-182 SGT B-2 Flow 2 3 4 Js def B NA Field Team Results Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 29 of 65 I/Noi.fri'Niilne"* lterfiate lhdicationiandltc)

TDescriptioif ompensato,y Measilre:s SOFT-TECH-ECN B NA Form 24075 SOFT-TECH- eDNA Application EDNA Service NA B eDNA Real Time Client or POIS Dose Software Emergency Dose RA1.1 I 1 I 2 I 3 I 4 I 5 I def I Projection System RS1.1 Redundant dose projection B

station or Table 3 of PPM 13.1.1 RG1.1 SPTM-Tl-5 Suppression Pool Avg I1 I2 I3 I Temp MA3.1 B MU3.1 Redundant instrument

• HCTL

• PC Potential Loss MS6.1 SW-RIS-604 RHR-HX-1A/B Outlet I 2 I 3 I 4 I 5 I def I Perform grab sample every 12 SW-RIS-605 Radiation RU1.1 hours and analyze for See PPM 1.10.1 A1 (0.1-1,000,000 CPS) radioactivity TDAS-SOFT- Plant Data Information POIS System (Control I 1 I 2 I 3 I 4 I 5 I def I B NA Control Room instruments Room)

TEA-FT-13 TG Building Elevated I 2 I 3 I 4 15 1def I Release Air Flow NA Alternate Method per PPM A1 Refer to ODCM 6.1.2.D 13.8.1 TEA-RIS-13 Turbine Building Refer to PPM 13.8.1 Exhaust I 1 I 2 I 3 I 4 I 5 I def I RG1.1 Attachment 7 .6 for converting frisker readings RS1.1 Notify Chemistry to set up (in cpm) to equivalent A1 RA1.1 Sample Cart (Frisker in a Brick) TEA-RIS-13 readings (in per 12.5.358 µCi/cc).

RU1.1 Sample cart is an Alternate METHOD.

Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 30 of 65 e 1nd1catiori a.Hal&

omi;ie'nsatotiMeasure:J TSC-CP-RAD/1 Notify the TSC EP Planner.

Monitoring Rack Establish periodic radiological A2 NA monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

TSC Mechanical Notify the TSC EP Planner.

TSC-FN-21 Equipment Room 12 13 14 15 I def I Establish periodic radiological 1-3 CFM Sample A2 NA monitoring of the TSC. If a FAZ Pump occurs, consider relocating the TSC to the Main Control Room.

TSC Airborne Notify the TSC EP Planner.

TSC-RE-1A Radiation Detector - 12 13 14 15 I def I Establish periodic radiological Particulate A2 NA monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

TSC Airborne TSC-RE-18 Radiation Detector - 12 13 14 15 I def I Notify the TSC EP Planner.

Establish periodic radiological Iodine A2 NA monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

TSC-RE-1C TSC Airborne Notify the TSC EP Planner.

Radiation Detector - 12 13 14 15 I def I Establish periodic radiological A2 Noble Gas NA monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

TSC Airborne Notify the TSC EP Planner.

TSC-R1S-1A Radiation Monitor - 12 13 14 15 I def I Establish periodic radiological A2 Particulate NA monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

TSC-RIS-18 TSC Airborne Notify the TSC EP Planner.

Radiation Monitor -

1 12 13 14 15 I def 1 Establish periodic radiological A2 Iodine NA monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: NIA

Title:

EP Equipment Page: 31 of 65 easure TSC Airborne Notify the TSC EP Planner.

Radiation Monitor - Establish periodic radiological A2 NA Noble Gas monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

TSC-RR-1 TSC Airborne Notify the TSC EP Planner.

Radiation Monitor 12 I 3 I 4 I 5 I def I Establish periodic radiological Recorder A2 NA monitoring of the TSC. If a FAZ occurs, consider relocating the TSC to the Main Control Room.

TSW-RIS-5 Ratemeter-Plant Serv Water Sample Rack 1 I 2 I 3 I 4 I 5 I def I A1 RU1.1 Perform grab sample once per See PPM 1.10.1 shift TSW-SR-34 WEA-RIS-14 Rad Waste Building Refer to PPM 13.8.1 Exhaust I 2 I 3 I 4 I 5 I def I RG1.1 Attachment 7.6 for converting frisker readings RS1.1 Notify Chemistry to set up (in cpm) to equivalent A1 Sample Cart (Frisker in a brick)

RA1.1 WEA-RIS-14 readings (in per PPM 12.5.37.

µCi/cc).

RU1.1 Sample cart is an Alternate METHOD.

WEA-SUM-1 RW Bldg Exhaust Stack Air Flow 1 I 2 I 3 I 4 I 5 I def A1 NA Refer to ODCM 6.1.2. D Alternate Method END Attachment 8.1, EP Related Important Equipment by EPN

• Number: 13.14.11

• IUse Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 32 of 65 Emergency Response See PPM 1.10.1 Facility Function - Alternate Emergency Preparedness is As determined by EOF A2 NA to be informed prior to ANY Emergency Preparedness maintenance activity in this center.

• • • • -*..--*****-*-***-**--*---**--*--*-------t-.=====;==;===;====r==;====,t----- ---**-**---------*-***----*-----------

~~::r~jection I 2 I 3 I 4 I 5 I def I RG1.1 Stand alone software installed on selected RS1.1 Perform dose projections B individual computers RA1.1 in other centers EAL Table 3 Inter Facility Communications 1 2 1 3 1 4 1 s I def 1 Dedicated X300 and x500 Circuits B NA Dial-Up Line Ring down Line B NA Commercial Line

- - - - - - - - + - - - - - - * * - -------****---*-*-*-*-**-***--***-**--- ***- *****-*-**--*--*****-*-***-*-***-**-----*-**-**---*-*****--*****-

Commercial Line No satellite phone assigned B NA Satellite Phone to the Alternate EOF. One available in the JIG.

---

*-------***----- **------ - - - - - * - - * - - - - - - - - - l FAX B NA Scan documents to email Field Team Radios Field Teams would be B NA Cell Phone dispatched to the plant to retrieve Field Team vehicles Ventilation Initiate work request 1 2 1 3 1 4 1 5 1 def 1 Contact EOF Planner B NA Establish alternate cooling/heating as needed Attachment 8.2, EP Related Important Equipment by Function

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 33 of 65 Emergency Response Facility Function - Control B NA Remote Shutdown Room Room AC Power 1 2 1 3 1 4 1 s I def 1 B NA PPM 5.6.1 - Station Blackout Dose Projection Standalone software Software 1 2 1 3 1 4 1 s I def 1 RG1.1 installed on selected RS1.1 Perform dose projections B individual computers RA1.1 in other centers EAL Table 3

~~8:i~~~iil~~tions 1 I 2 I 3 I 4 I 5 I def I ----------+-------*

Dedicated X300 ABN-Communication B NA and x500 Circuits Dial-Up Line t-------t-------1---~---*----+-----------1 Dial-Up Line ABN-Communication B NA Commercial Line

---

*------* ----1------*

Ring down Line ABN-Communication B NA

-*-----*-**--*-**--**--..-*-..---*--- Commercial Line Commercial Line ABN-Communication Satellite phone stored in B NA Satellite Phone TSC

>------------- --------t-*---*----***----***------------------**------

FAX B NA Scan documents to email Repair Team

~------*-----------1---

B NA Plant phone system Radios Ventilation B NA ABN-HVAC Attachment 8.2, EP Related Important Equipment by Function

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 34 of 65

• • ~~'**tt/ 1<Alternate lridfoatiori'.and ****

... tor Compensatory~ .

Measuret**

Emergency Response See PPM 1.10.1 Facility Function - EOF Emergency Preparedness is B NA Alternate EOF to be informed prior to ANY maintenance activity in this center.

AC Power 115 KV Line supplies power 12 I 3 I 4 I5 I def I to Kootenai - EOF.

B NA Diesel (Bldg 34 DG) Diesel (Bldg 34 DG) supplies EOF Backup Power for RED Outlets and Emergency Lighting Standalone software Dose Projection RG1.1 installed on selected Software RS1.1 Perform dose projections B individual computers RA1.1 in other centers EAL Table 3 Health Physics Center 11 I 2 I 3 I 4 I 5 I def I Use decontamination Decontamination . . . . . . . B NA facilities located in-plant Shower 1-------+-----i~(~R_W_4_87]_~~<:!__atENOC Health Physics Center Contact EOF Planner ERO to implement alternate Receiving Area B NA location as defined by Initiate work request 13.11.7 Inter Facility Communications Dedicated X300 B NA Dial-Up Line and x500 Circuits 1----*

Dial-Up Line B NA Commercial Line Ring down Line B NA Commercial Line Commercial Line B NA Satellite Phone

,-------~------, ------*--+------ - - - - - - - - - - - - - - - - - - - - i FAX B NA Scan documents to email

---

11-------------

Field Team Radios B NA Cell Phone Attachment 8.2, EP Related Important Equipment by Function

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 35 of 65

• Noun NarrieTDescrjption ..

I orCorripensatory:

• • >Measure

• NA Ventilation Contact facilities and (EOF) confirm Kootenai HVAC EOF confirms habitability B NA upon reporting and relocates Establish alternate if necessary.

cooling/heating as needed Air Handling Unit 1 Verify normal Kootenai AHU-1 HVAC available Backup to normal Kootenai A2 NA HVAC Relocate to Alternate EOF HVAC Test performed

• - - - - - - * - - * - - - EOF if. neces._s_ary~--- quarterly by MWO OOGCF3 1-------*---*----l Fan SF-3 Verify normal Kootenai HVAC available A2 NA Supply Fan 3 Relocate to Alternate

---**-*----* -***------** EOF if necessary --*-*-**-**-*- *-*-*---*---*--**--**-------

Radiation Monitor, Confirm no release in Intake Air progress Only required if center is A2 NA activated.

Shift to filtered air if

• ----+-re_l_ea_s_e in__,_p_ro~g~r_es_s_ _ _ t - - - - - - - - - - - - - 1 Radiation Monitor, Confirm no release in Return Air progress Only required if center is A2 NA activated.

Shift to filtered air if release in progress Attachment 8.2, EP Related Important Equipment by Function

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: NIA

Title:

EP Equipment Page: 36 of 65 rnat~lridicationarid :

...... tor Compensatory * * * *

• Measure Emergency Response See PPM 1.10.1 Facility Function - JIG Notify JIG Planner Emergency Preparedness is A2 NA As determined by to be informed prior to ANY Emergency Preparedness maintenance activity in this center.

AC Power 1 12 13 I 4 I 5 I def I B NA Diesel backup power Inter Facility Communications I 2 I 3 I 4 I 5 I def I

---

***-------*-  !--------+------- -----*-*--

Dedicated X300 B NA Dial-Up Line and x500 Circuits Dial-Up Line B NA Commercial Line 1--------+---------------------1---------*-------<

Ring down Line B NA Commercial Line Commercial Line B NA Satellite Phone

---

+------r----------t------------~

FAX B NA Scan documents to email Ventilation Initiate work request 11 I 2 I 3 I 4 I 5 I def I Contact JIG Planner B NA Establish alternate cooling/heating as needed News Conference Capability - I 2 I 3 I 4 I 5 I def I B NA MPF Auditorium Skamania Room Walkley Rm Attachment 8.2, EP Related Important Equipment by Function

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 37 of 65

' Alternate lndictttic:frland i

).forCorntiertsatoot ..

• 'Measure NA Emergency Response Notify OSC Planner See PPM 1.10.1 (OSC) Facility Function - OSC Emergency Preparedness is A2 NA Relocate the OSC as per to be informed prior to ANY "OSC Manager Checklist" maintenance activity in this (Form 26522) center.

AC Power Relocate the OSC as per 12 13 I 4 I 5 I def I B NA "OSC Manager Checklist" (Form 26522)

Inter Facility B NA Communications I 1 1213 14151 def I Dedicated X300 and x500 Circuits Dial-Up Line Dial-Up Line Commercial Line Ring down Line Commercial Line Commercial Line OSC has no Emergency None Response Function that

__i:eq~i!"es commercial lines.

FAX Scan documents to email Repair Team Plant phones Radios TSC Building Ventilation I1 1 2

1 3

1 4

1 5

1 def I Initiate work request Contact OSC Planner OSC confirms habitability B NA upon reporting and relocates Establish alternate if necessary cooling/heating as needed Attachment 8.2, EP Related Important Equipment by Function

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 38 of 65

.!

!:!.!::*" i:').J :;::A1terriatcflndfoatioh'.ana****

• • • • • • • • -***-**1*or**.Compe*nsafory.

. Measure ...

See PPM 1.10.1 Notify OSC Planner Emergency Preparedness is B NA Relocate the TSC as per to be informed prior to ANY "TSC Manager Checklist" maintenance activity in this (Form 26506) center.

AC Power Relocate the TSC as per 1 I 2 I 3 I 4 I5 I def I B NA "TSC Manager Checklist" See Attachment 8.1 (Form 26506)

Dose Projection Standalone software Software 12 I 3 I 4 I 5 I def I RG1.1 installed on selected RS1.1 Perform dose projections B individual computers RA1.1 in other centers EAL Table 3 Inter Facility B NA Communications 1 I 2 I 3 I4 I5 I def I Dedicated X300 and x500 Circuits Dial-Up Line Dial-Up Line Commercial Line Ring down Line Commercial Line Commercial Line Satellite Phone

• ---------*--- -*-*-*------**---1 FAX Scan documents to email Ventilation Initiate work request 1 2 1 3 14 I 5 I def I TSC confirms habitability Contact TSC Planner upon reporting and relocates B NA if necessary Establish alternate cooling/heating as See Attachment 8.1 needed Attachment 8.2, EP Related Important Equipment by Function

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 39 of 65

---

! ------------t------------------------------------------------- -----------------------------

Auto-Dialer Required only when B NA Planner Call-Tree Radio Paging System unavailable Normal Power Also powers company B NA Backup Generator phone system Backup Generator Required only when Confirm Normal Power normal power unavailable B NA Contact Emergency Backup generator known as Preparedness TEL-COM-GEN,30-209, or DG-2(TEC)

Cellular Phone B NA Commercial Line

• Commercial Line B NA Cellular Phone Radio Paging Manual Activation Radio paging is the B NA OR primary method of ERO notification.

Auto Dialer 1------------------------------

Public Address Make PA Announcement B NA from other center.

NA NRG Notification Function B NA See PPM 1.10.1 (NRG Notification) ENS Line Commercial NRG HOO Line HPN Line Commercial Line EROS Verbally via ENS call FTS 2001 Commercial Line Attachment 8.2, EP Related Important Equipment by Function

• Number: 13.14.11

• J Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 40 of 65 t*ii****************** **,< ror Compensatory Measure NA Offsite Notification Function B See PPM 1.10.1 (Offsite Notification)

CRASH Line Dial-Up Line 1------*-----*t---*

Dial-Up Line Commercial Line Commercial Line Alternate phone switch (377, 372, 375)

Normal Power Also powers company Backup Generator

.... __phone system Backup Generator Confirm Normal Power Backup generator known as TEL-COM-GEN,30-209, or DG-2(TEC)

Contact Emergency Required only when Preparedness normal power unavailable Radio Alternate channels Commercial Line LERN Commercial Line SCAN Commercial Line FAX CRASH Line Commercial Fax Attachment 8.2, EP Related Important Equipment by Function

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 41 of 65 tor Corripehsafory

.Measure NA Site Evacuation Notification See PPM 1.10.1 Function CGS Site Evacuation Public Address or Area Siren System Sweep B NA HNES Units Area Sweep NA Pulic Alert & Notification System (ANS) Function 12 13 I 4 I s I def I See PPM 1.10.1 Sirens For siren system loss: For single siren loss:

(Primary Public ANS) Confirm functionality of Contact IS to confirm ETNS system. functionality of adjacent sirens Emergency Telephone Confirm functionality of Contact County EM Staff Notification System Public ANS Siren system. and ETNS vendor for (ETNS)

B NA resolution as needed (Backup Public ANS)

Tone Alert Radios KONA AM/FM Radio **TARs are in very limited (TARs)** KORD FM Radio use within the CGS 10-Mile Emergency Planning Zone.

See CGS ANS Design Report for details.

END Attachment 8.2, EP Related Important Equipment by Function

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 42 of 65

.*A1t~fnat~*1~aid~ti~~*~ria******

... Eauipiri~"!.* *ror*compehsafory <

. Cate!JOfY .** > ,,.,........ Measure.) .. *********** . L<>>L NA Computers B NA Contact IS (TSC) 1----------------l Chemistry/ Effluent Engineer 1 Engineer 2 Engineer 3 Information Coordinator NRC Liaison NRC Computer Print station w/ large PORTAL J or hard copies CPU and Monitor monitor from TSC Library NA Printers/Copiers B NA Contact IS (TSC) I 2 13 I 4 I 5 I def I Print Station Plotter Printer, Engineering Area Lan Printer/Copy Contact Vendor Also for OSC Machine NA Fax Machines B NA Contact IS Also for OSC (TSC)

Dedicated Commercial Attachment 8.3, EP Related Important Equipment - TSC

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 43 of 65 NA Headset, Information B Contact IS (TSC) Coordinator NA Telephone B NA Contact IS Inventoried / tested by (TSC) ------------*----------------------------------l 1 I 2 I 3 I 4 I s I def I Telecom TSC Manager Operations Manager Radiation Protection Manager Technical Manager PlanUNRC Liaison Administrative Manager Chemistry/Effluent Manager Maintenance Manager Plant Technical Staff Plant Technical Staff 1--------------------

Plant Technical Staff Plant Technical Staff Security Liaison Adm in Support Staff Computer Engineer Attachment 8.3, EP Related Important Equipment - TSC

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 44 of 65 Altei'riatei!fhaicii'tion:an rorco NRC Emergency B NA Contact IS Headset and phone Notification System (ENS)

NA NRC Communications B NA Contact IS Inventoried / tested by (TSC) 1 2 1 3 I 4 I 5 I def I Telecom NRC Health Physics Network (HPN)

NRC Management Counterpart Link (MCL)

NRC Rad Safety Coordinator NRC Reactor Safety Counterpart Link (RSCL)

NRC Reactor Safety Ops Coordinator NRC Reactor Safety Specialist NRC Senior Resident Inspector

-****-------*-**-*------*--*----I NRC Protective Measures Counterpart Link (PMCL)

Attachment 8.3, EP Related Important Equipment-TSC

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 45 of 65 NA (TSC)

Radios

---

i1 1 2 I 3 I 4 I s I def I 8 NA Contact IS Inventoried and tested by Telecom Base Station (x2)

There are 2 more base On~Site Channels stations for OSC Area Wide channels (total of 4)

NA Flat Screen Monitors 8 NA White Board (TSC) 1 2 1 3 J 4 J s J def I END Attachment 8.3, EP Related Important Equipment - TSC

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 46 of 65 Alternate'. lndi6ation and:;.

l or Compensatory

• • • • • • • • /Measure NA Computers B NA Contact IS Tested by EP, (OSC) Maintained by IT HP Lead Repair Team Coordinator/ Manager Electrical Craft Lead l&C Craft Lead Mechanical Lead NA (OSC)

Repair Team Coordinator Headset I 2 I 3 I 4 I 5 I def I B NA Contact IS NA Telephones B NA Contact IS Inventoried and tested by (OSC) f--------------4 11 I 2 I 3 I 4 I 5 I def I Telecom Manager HP Lead Repair Team Coordinator Team Tracker Electrical Craft Lead l&C Craft Lead Mechanical Craft Lead Craft Area Attachment 8.4, EP Related Important Equipment- OSC

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 47 of 65

,r.'.'7=,

=*. .. .. **-" ~ ,;:;,-r,*c-,,.*.;:;,.;;-.*.,"""""'"'""'".T"** . . . . .. . ,.

.!

• • • • ::Ji:!:: r:l*iA1t~'~ti~tl1hdititiJ~Cihc:tf I ofCOmpensatory* ., .

'Measure.*******

NA Radios B NA Contact IS Inventoried and tested by (OSC) Telecom Base Station (x2)

There are 2 more base On-Site Channels stations for TSC (total of 4)

Area Wide channels Repair Team hand held (x6)

END Attachment 8.4, EP Related Important Equipment- OSC

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 48 of 65 EP RELATED IMPORTANT EQUIPMENT-EOF (BUILDING 34)

• • • • • • • • i INPO 10:001 :s Eg1.1iprrierit(.

• • .**category;:

B NA Contact IS Tested by EP, Maintained by IT Asst EOF Manager DOH Dose Projection Energy Northwest Dose Projection EOF PIO Information Coordinator (x2)

INPO Network POIS Analyst Radiation Detection Engineer Telecom Manager NA Emergency Diesel A2 Confirm 115 kV available DG Load Test performed (EOF) Generator 1 2 1 3 I 4 I 5 I def I Relocate to Alternate monthly by EOF. MWO#OOGBKO DG Switch Test performed quarterly by MWO#OOGBK2 Attachment 8.5, EP Related Important Equipment - EOF (Building 34)

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 49 of 65

_:\/,.:.:*,/':\:: :*;::,;;;::=.:*:::=:

...* ,INP010;007. ****t-.1t~t~~tJ iriil1E~tia~***~;,d*

iEguiament< *** forCom12ehsato~.**

Category* *Measure r"***.*.* * *

• Fax 1 3 4 5 B NA Contact IS Inventoried / tested by Telecom - EP CNF Dialogic Incoming Outgoing NA Health Physics Center B NA Kit at Alternate EOF Inventoried/tested by (EOF) Vehicle Field Team Air I2 I 3 4 5 I def I Contact HP to provide Radiation Protection Sample Kit (x3) additi~nal equipment quarterly- MWO 01050707 NA Health Physics Center B NA Kit at Alternate EOF Inventoried/tested by (EOF) Vehicle Field Team Field 1

I2 I3 4 5 I def I Contact HP to provide Radiation Protection Sample Kit (x3) additional equipment quarterly- MWO 01050707 NA Health Physics Center B NA Kit at Alternate EOF Inventoried/tested by (EOF) Vehicle Field Team I2 I3 4 5 I def I Contact HP to provide Radiation Protection Instrumentation Kit (x3) additional equipment* quarterly- MWO 01050707 NA Health Physics Center B NA Kit at Alternate EOF Inventoried/tested by (EOF) Vehicle Field Team I2 I 3 4 5 I def I Contact HP to provide Radiation Protection Protective Clothing Kit (x3) additional equipment quarterly - MWO 01050707 NA Printers B NA Contact IS Tested by EP (EOF) --------*-- I2 I 3 4 5 I def I HP Dose Projection INPO Network Coordinator Laser24 Laser147 NA Radiation Monitor, 4 5 I B NA Portable Monitor Tested I Calibrated by RP (EOF) Victoreen VAMP I2 I3 I def Attachment 8.5, EP Related Important Equipment- EOF (Building 34)

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 50 of 65 NA B NA Contact IS Inventoried and maintained (EOF) by Telecom Field Team Console EOF Radio inventory Field Team Monitoring Field Team Vehicle EN1 Field Team Vehicle EN2 Field Team Vehicle EN3 Portable Field Team (x6)

NOAA Weather WA State Field Team Console Security Area Wide Call sign KZl509 DOE Patrol Call sign Station 51 DOE Safety Console Call sign Station 51 Law Enforcement Call sign KOM785 Radio Network Attachment 8.5, EP Related Important Equipment - EOF (Building 34)

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 51 of 65 3 4 5 Admin Support Admin Support Assistant Manager Benton County Emergency Management BPA Representative FEMA Representative Decon Showers Dept. of Energy Rep resentatiye Dept. of Health Field Team Coordinator Dept. of Health Protective Action Decision Group Liaison Dose Projection HP Dose Projection HP Attachment 8.5, EP Related Important Equipment - EOF (Building 34)

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 52 of 65

• 1NP010;001 \ ***li.11~:friatJ**11iait~1,bh**a11d**i Equipment ** Ior Compensatory F **

Categofy\*

• Measure<**.

Engineering General 3 4 B NA Contact IS Inventoried/ tested by Area Telecom Engineering Manager Engineering Support Field Team Coordinator Field Team Dispatcher Franklin County Emergency Management Health Physics Center INPO Network Coordinator Instrument Calibration Lab Manager Manpower Scheduler NRC Deputy Site Team Director NRC Deputy Protective Measures Branch Leader

• -***-**-*-*-**--*--*-----*--*~---------~-----~----~--------~-----------'

Attachment 8.5, EP Related Important Equipment - EOF (Building 34)

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 53 of 65 3 4 5 Inventoried/ tested by Telecom NRG Dose Assessor NRG Dose Assessor Modem NRG ENS NRG Information Technology Specialist NRG Liaison Leader NRG Management Counterpart Link NRG Protective Measures Counterpart Link NRG Protective Measures Branch Leader NRG Chronology Documentation Communicator NRG Deputy Technical Assessment Branch Leader

.Attachment 8.5, EP Related Important Equipment - EOF (Building 34)

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 54 of 65 NA NRC Deputy B 3 4 5 (EOF) Technical Assessment Branch Leader NRC Deputy Technical Assessment Branch Leader Modem NRC Reactor Safety Counterpart Link (RSCL)

NRC Response Branch Leader NRC Response Branch Leader Modem NRC Site Team Director NRC Technical Assessment Specialist NRC Technical Assessment Branch Leader Oregon State Liaison Oregon Radiological Dose Analyst PDIS Analyst Attachment 8.5, EP Related Important Equipment- EOF (Building 34)

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: NIA

Title:

EP Equipment Page: 55 of 65 Public Information B NA Contact IS Inventoried / tested by Officer (PIO) Telecom Radiation Detections Systems Engineer Radiological Emergency Manager (REM)

Respiratory Testing Area Secretary Secretary Security Manager Site Support Manager State/County Liaison Technical Liaison Telecom Manager Washington Dept. of Agriculture Washington State Emergency Management Liaison Washington State DOH Liaison

- - - - ~ - - - - - - - - ~ - - - -..---*------------------**- --------------------~---..-----

Attachment 8.5, EP Related Important Equipment - EOF (Building 34)

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: NIA

Title:

EP Equipment Page: 56 of 65 B NA Acquire other ENW Maintained by CMS -

• ---*--*--*------*-******-*- vehicle EP verifies operable on a Field Team EN1 Contact Vehicle monthly basis

---

*-*-*--------***---..*..---**-*-***- Maintenance Field Team EN2 Contact WA to request Field Team EN3 DOH supplement NA Whelen Siren Console, Site B NA Activate from sec Inventoried/Tested by (EOF) Evacuation Siren System 12 I3 4 5 I def I Telecom NA Wireless CRASH phone B NA Contact IS Inventoried/Tested by (EOF) system 12 I3 4 Is I def I Telecom and EP NA Overhead Projector (x2) B NA Contact IS (EOF) 12 I3 4 5 I def I NA Flat Screen Monitors B NA Contact IS (EOF) 12 13 I 415 I def I END Attachment 8.5, EP Related Important Equipment - EOF (Building 34)

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 57 of 65 EP RELATED IMPORTANT EQUIPMENT-ALTERNATE EOF B.5.b Radios Cellular phones Field Team Base Cellular phones Station NA Fax B Inventoried/tested by (Alt. EOF) 1 2 I 3 1 4 I 5 J def I Telecommunications NA Health Physics Center B NA Kit at EOF Inventoried/tested by (Alt. EOF) Vehicle Field Team Air Contact HP to provide Radiation Protection Sampling Kit additional equipment quarterly- MWO 01050707 Health Physics Center Vehicle Field Team Field Sample Kit Health Physics Center Vehicle Field Team Instrumentation Kit NA Telephones B NA Contact IS Inventoried I tested by (Alt. EOF) 1 2 I 3 1 4 I s J def I Telecom Assistant Manager Benton County Commissioner Dept. of Energy Field Team Coordinator Dept. of Energy Field Team Dispatcher

• -*-**-**-******----.l----*-*-----*-----*--------- -*-----------*-------*- --*----*------*-----* *---*--*-*-- ----*--*-*-*-*------------- *-*----------*---------*-*-*-

Attachment 8.6, EP Related Important Equipment - Alternate EOF

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 58 of 65

• 1NP010Looz>.! ,. ltern~te lhiiid~ti6ri ~nd:

Etji.Jipmenf i I or Compens'atorv**

Category ** **Measur~<*

B NA Contact IS Inventoried/ tested by Telecom Engineering Manager Manager Secretary Field Team Coordinator Field Team Dispatch Franklin County Commissioner Manpower Scheduler 1------------

NRC/FEMA NRC Protective Measures Branch Lead NRC Reactor Safety Counterpart Link (RSCL) Communicator NRC Response Coordination Leader NRC Site Team Director

---

*----------~---------~-----~------**

Attachment 8.6, EP Related Important Equipment - Alternate EOF

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 59 of 65 Assessment Branch Leader Oregon State Liaison REM Site Support Manager Security Manager Technical Liaison Telecom Manager Washington State Representative Washington Department of Health NA (Alt. EOF)

Headset I 2 I 3 I 4 I s I def I B NA Contact IS Controller and Information Coordinator NA Computer, Dose Projection B NA Contact IS (Alt. EOF)

END Attachment 8.6, EP Related Important Equipment -Alternate EOF

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 60 of 65 NA AM Tuner B NA FM Tuner (JIC) Internet NA Computers B NA Contact IS Tested by EP, (JIC) 1 I 2 I 3 I 4 I s I def I Maintained by IT Alternate EOF/ENS AV Walkley Room Benton County PIO Franklin County PIO HP Spokesperson Internet News Monitor JIC Secretary News Release Editor NRC PIL Oregon PIO Phone Team (x8)

Rumor Control Technical Spokesperson) 1----*--*-*****--------------

Washington PIO)

Attachment 8.7, EP Related Important Equipment- JIC

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 61 of 65 NA DISH Network Receiver (x2) B NA Contact Vendor (JIG)

NA Fax Inventoried/Tested by (JIG)

Telecom Commercial (x2) B NA Contact IS NA Dedicated (x2) 1 12 3 4 is def I B Contact IS NA Flat panel displays (5) 4 is B NA Contact IS JIG, Phone Team Room (JIG) I2 3 def I and MPF Lobby NA FM Tuner 4 is B NA AM Tuner (JIG) 12 3 def I Internet NA Headset 4 is B NA Contact IS (JIG) --*-----*---*-*-*----! 12 3 def I Phone Team, (x8)

Spokesperson NA Printers B NA Contact Vendor (JIG)

PIO Work Area Rumor Control Work Area NA Projection Booth AV B NA Contact IS Prox card and booth key (JIG) Equipment and DVR located in Support Manager's center desk drawer Attachment 8.7, EP Related Important Equipment- JIC

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 62 of 65 NA Telephones B NA Contact IS Inventoried/Tested by (JIC) Telecom AudioNisual Benton County Representative Distribution Team Supervisor DOE-RL Representative FEMA Representative Franklin County Representative HP Spokesperson Information Manager Manager Media Phone Team News Release Editor NRC Representative Oregon State Representative Public Phone Team Receptionist

• - - - - . L - - - - - - * * * - * * *..-----*--- --*---*----***-----*--* ---*--**--*--**--*--**--*-- ---****-*-*----

Attachment 8.7, EP Related Important Equipment-JIC

• Number: 13.14.11

• I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 63 of 65 NA (JIC) ***------*--*-*-*****--*----

Support Manager Technical Spokesperson Washington State Representative EMO Washington Department of Agriculture Washington Department of Health NA Televisions, broadcast (x2) B NA White Board (JIC) 12 13 14 I s I def I NA Video B NA Contact IS (JIC) I 2 I 3 I 4 I s I def I JIC Projector Walkley Room Projector END Attachment 8.7, EP Related Important Equipment- JIC

• Number: 13.14.11

• Use Category: REFERENCE 1 - - - - - - - - - - - - - - - - - - - - ' - - - - - - - - - - - - - - - l Minor Major Rev: 013 Rev: N/A

Title:

EP Equipment Page: 64 of 65 UNPLANNED LOSS OF EQUIPMENT Generate appropriate Unplanned deficiency tracking maintenance or loss documents and of function prioritize to restore function YES (CatA) Implement YES Evaluate reportability compensatory requirements measure(s)

NO Review with mgmt. and determine actions Identify redundant or backup indication or method Track OOS equipment

~ - - - - - - and return to service date Communicate need for Reclassify equipment heightened as Cat A. Reprioritize management work accordingly. awareness and deficiency tracking

• A function requires more time to Perform work as perform or cannot be performed.

necessary END Attachment 8.8, Unplanned Loss of Equipment

Number: 13.14.11 I Use Category: REFERENCE Major Rev: 013 Minor Rev: N/A

Title:

EP Equipment Page: 65 of 65 PLANNED LOSS OF EQUIPMENT Evaluate maintenance

~ - -.. activity or planned loss of function NO YES (Cat A) YES Is the compensatory Evaluate reportability measure available? requirements NO NO Review with mgmt. and (CatB) determine actions Implement compensatory NO measure(s) prior to removal from service

• YES

' - - - - - - - - - - - Notify ERO as necessary 1 4 - - - - - - ~

Perform work as scheduled

• A function requires more time to perform or cannot be performed.

END

• Attachment 8.9, Planned Loss of Equipment