Text

Proposed Upgraded Emergency Action Levels
	ML040630161
Person / Time
Site:	Arkansas Nuclear
Issue date:	02/27/2004
From:	James D; Entergy Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	0CAN020407
	Download: ML040630161 (269)
	v • d • e

Entergy Operations, Inc.

Entergy Ene gy JRussellville, 1448 S.R. 333 AR 72802 Tel 501 858 5000 OCAN020407 February 27, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Proposed Upgraded Emergency Action Levels Arkansas Nuclear One - Units 1 and 2 Docket Nos. 50-313 and 50-368 License Nos. DPR-51 and NPF-6

Dear Sir or Madam:

Upgraded Emergency Action Levels (EALs) for Arkansas Nuclear One (ANO), Units 1 and 2 are enclosed for NRC staff review and approval as required by 10CFR50 Appendix E, IV.B. These new EALs were written using the methodology outlined in NEI 99-01, Methodology for Development of Emergency Action Levels, Revision 4, January 2003. NEI 99-01 has been endorsed by the NRC Staff in Regulatory Guide 1.101, Revision 4, July 2003, Emergency Planning and Preparedness for Nuclear Power Reactors and in NRC Regulatory Issue Summary 2003-18, October 8, 2003, Methodology for Development of Emergency Action Levels. The proposed changes to the ANO EALs have been reviewed and approved by the Onsite Safety Review Committee. Additionally, agreement on the proposed EALs has been obtained from the State of Arkansas and local governmental authorities.

The guidance contained in NEI 99-01 was found to be acceptable to the NRC staff as an alternative method for development of EALs to that described in Appendix 1 to NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants and in NUMARC/NESP-007, Methodology for Development of Emergency Action Levels.

Plant specific information is attached in the following order:

Current Emergency Plan Mark-Up

Proposed Emergency Plan Pages - Changes Incorporated

Proposed EALs

Arkansas Nuclear One Deviations and Differences from the NEI 99-01, Revision 4 Emergency Action Levels A ?'

OCAN020407 Page 2 This correspondence contains no new commitments. If you have any questions regarding this submittal, please contact Mr. Robert Holeyfield, Manager, Emergency Planning at 479-858-4995.

Sincerely,

/Dale E. ames

( Direct r, Nuclear Safety Assurance (Acting)

DEJ/fpv Attachments:

1. Current Emergency Plan Mark-Up

2. Proposed Emergency Plan Changes - Changes Incorporated

3. Proposed EALs

4. Arkansas Nuclear One Deviations and Differences from the NEI 99-01, Revision 4 Emergency Action Levels

OCAN020407 Page 3 cc: Dr. Bruce S. Mallett Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P.O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Thomas Alexion Mail Stop 0-7 D1 Washington, DC 20555-0001

Attachment I OCAN020407 Current Emergency Plan Mark-Up

UNiT ONE EMERGENCY G[ASS INMA-ATNG CONDMTONS NOTIFICGAT)IN OF UNUSUAL EvE 4 -. ^A A -s

!i-'i:~i~' kl bI I -

Lii l AE -

t:tI'-Lv

~

a) RCS activity indicates >0.1% fue eI ladding failUr.

2. REACTOR COOLANT SYSTEM LEAKAGE a) Suleakage EN > v pS unidetfe er RG Pressure Beundary leakg er>

25 gpm identified RCS lealcage

&-. SEGDRYSYTMEET a) Unconitrollcd GT-SG depressurizatien rcsulting in MSLI aetuation.

b) OTSG Tube Lcakagc > Tcch. Spec. Limits

1. ELETRICAL POWvvER FALRV16ES a) Degraded pewer T~l I r% r% AI "--. I lf-.l 5-. nlA ImAbflUdbU1LAL trrhtL~t~1 a) Projcctd or mcasured activity at thc sitc boundary, avcaged ovcr onc hour, is g;rater than or cqual to .05 mrem/hr TEDE or .15 mrem/hr Child Thyroid CDE or liquid r-adielegleal effluents exeeed ODCM limitts.

6.SAFFEMY- SSTMFNG4GN a) Dcviation from Tcchniael Specification action ztatemcnt when required to shutdown or cooldown oedcviations pursuant to 10CFR5O.51(x).

b) Loss of dee assess ment capabilities.

e) Loss of cowmlunications.

Revision 29 D-4

TABLE D-1 (Gentinu^d) 7 7

tfA -A nrf

)'-

It fA I

TI III ratl _ .

nrnr II'1 I IIII!..

IIIrfl t a) Scurity threat onsitc but outsidc thc Protceted Arca Sccurity Fcnce (c.g.

attempted cntry or sabotagc which has been stopped outsidc the security b) Fire or explesien ensiti.

or flammable gas release.

8. NATURAL EVENTS a) Tornado, flood, les: of Dardancilc Rcscrvoir, carthquakc.

9. MISCELLANEOUS EVENTS a) Other plant conditions cxist that warrant incrcased awarencss on thc part e thc epcrating staff and state ander/or la! _ffItF autheritisoer involve ether than a normal controllcd shutdown.

Revision 29 D-5

TABLE D 1 (Gentinued)

ALERT I flfTKAAflV CVe=Rr'A B a %+/-I ar,' j I j I I-V I I MlE7KNrTC IF I 'l I .-

2. REAGT49R COOLXANT SYSTEM LEAK<AGE a) nRG leakage _ _e__l makeup capaity (50gpfn).

SECONDARY SYSTEM EVENTS a) OTSC less tubc ewithl_ 10_ltcam gm concurrcfrclcacongin r los o ofstc pewetr-

1. ELECT-RICALV P)WER FAILURES a) Station Blacl(out.

b) LOss of all vital DC power.

RADIOLOGICAL EFFLUENTS a) Projcctd or mcasured acivity at thc sitc boundary, avcvged over onc hour, is greater than orequal te .5 mrcn/hr TEDE or 1.5 mrmer/hr Child Thoyreld CDE or liquid radiological effluenits exeeed 10 times ODCM limits.

b) [Iiaf radiation,'awbrt rnlevels.

CAFT r-pl rTF.M Ft1 ih'.C~lI~

a) RPS failurc to cemplete an automatic trip.

b) Loss of control room annunciators.

c) Control Room cvacuation.

d) LOss of decay hcat removal capabilitics.

Revision 29 D-6

TABLE D 6I(Gontinued)

7. HAZARDS TOQ STATION OPERATION eutside of plant buildings.

b) Fire or cxplosion onsitc affecting onc train of any ES system.

c) Aircraft crash, missiles, toxic or flammabic gas affecting onc train of any ES system.

8. NA, RAL EVENTS a) Tornado, high winds, flood, lesS of Dardanclic Rcscrvoir, carthqual(o.

9. MISCELLANEOUS EVENTS a) Other plant conditions cxist that warrant prcautionary activation of thc Tcchnical Support Ccntcr and placing thc ncar sitc Emergency Operationz Facility and ether lcey cmergency personncl on stand by.

Revision 29 D-7

TABLE D 1 (Continued)

SEAREA EMERGENGY PPIMADY Y~TM X~ FV-NT' a) Corc damage indicated with an inadequatc core cooling condition.

. \

rnIi ontainment rad at an readines wnicri ind1cate L~uLA and > 3i% . fue laldnam

- A S P _ @ @_

JX .

- filluf%.

2. RE\A_:Q C-9LN SY-SVVTEMz zI L AEAUk a) nRGlakg >Ab5nenfll fnakeu r;:apac i~ty (50pfn) wt.h >dnO 1.0 fuelA el5ad f},u u W gz lFllweM;; y*, }.., l. u un b)-RGS leakage -:I.- H4P! capaeit-SECONDAR-Y SYSTM EV'ENTS a) OTSG tube rupture with primary to secondary leakage >normal makeup capacity (50gpmn) with ongoing steam release or less of eff-site peweF.

b)h -C tube leak > 1 gpmn with > 1% fuel cladding failue with ongoing stemrelease.

1.ELEGTRICAL POEVAWER FAILURE:S a) Blaeckout for more than 15 minutes-.

b) Loss of all vital DC power for more than 15 minutes.

5. RADIOLOGICAL EFFLUENTS a) Radielegieal effluents are grcatcr than or equal to 50 fnrem/hr TEDE or 150

_nes/hF Ghild Thy;Arnc _d GDEa heSt Beundar,,y_.

b) Spent fuel accident.

Revision 29 D-8

TABLE D I (Conitinued) tCAr'f-f%/ t'/e-rrPA r~I 1Fd11rfNlN a) RPS failurc to complctc a manual trip.

b) esof3 vsReem Gena e avvannunciateru withr a t~niefitW il1 pmr-es.

e) Contrel room cveauatien and control of shutdown systems not established in 15 mInues d) Dcgraded hot shutdown capability.

7. HAZARDS TO STATION OPERATION a) Ongoing security thrcat within plant buildings but not within thc Control Room or vital arcas.

b) Firc or cxplosion ensite affecting both trains of any ES system.

e) Alrcraft crash, missiles, toxic or flammable gas affecting both trains ef any ES -syste,l-

8. NA:hRAL E'V'ENTS a) Tornado, high winds, flood, loss of Dardancllc Rcscrvoir, carthqualc

9. MIC.GELIANEOUSI CEVENTS

a. Other plant conditions cxist that warrant activation of cmergency rcsponsc facilitics and monitoring tcams or a precautionary notifcation to thc public near the -site Revision 29 D-9

TABLE D-1 (CGntinued)

GENERAL EMERGENGY I I.

MMTYf A A "%I

.. _ . . .... .... M_ _ . _. .

%PV kA 1A IIf&

II*

a) Containmcnt radiation rcadings which indicatc LOCA and >50% fucl evcrheat.

b) Corc mclt.

c) Loss of or challengc to all thrcc fission product barricrs.

N/A

3. SECNDARY SYSTEM EVENTS N/A ELECFRICAL _. PO\ER FAILURES N/A

%A rTrI~ rf-T- AI -r'I I Irh Tt%

Ir I itjtLtjlat-Lt clt-bLctN I z-

!t.

a) Radiologic-al effluents are grcatcr than or equal to 250 mr~em/hr TEDE or 500 Fflem/hr Child-Thyroid CDE at the-site boundar.~Y.

VA tCVT% CNVOC-rrA rf1 1ki1nTh1 6;- _ . . _ . _ . _ . . . _ . . _ . _ .,

N/A 7, . HA7AfRS T= 'TAfNM nPERA rn a) Ongoing seeurity thrcat within the Control Room or vital arcas.

Revision 29 D-10

T-ABS6En {tinued) bAr In AI MIETK M 8&. lilts I I IlffEl rurlul I

N/A

. T. ~- I I A-~ r UP1r^ - -

9-. snuUf k bFruu t-l IUbtAd a) Plant conditions exist that make rclcasc of large amounts of radioactivity pessible.

Revision 29 D-11

TABLE D 2 UNlTTlW:PNO EMERGENCY CLASS INMATING CONDMONS NG4NFIGAYG ONUSUAL EVENTJF

-R. PRIMAR- SEM- EVENTS a) RGS activity indieates >0.1% fuel eladding falulwe . n flA f-rtf'

~I fl r^r ~

u WMj,~

Ah I I TrCNIPrkA I r! A /A

-PI ..j I 1I & I',.J rf~

a) nRG leakage^ > 1 n gp_ unidentified RG F Pressurverr Ben*

_~leakaget_ eF >

3. -SECGOENDARtY SY-STEM EVENTS a) Uneentrelled S/G depFessurizatien resulting in MSIS aetuatien.

b) S/G tube leak -A.- Teeh. Spe. Lmits A r2- 1-rn n A. flf A - V- A TO I Nnrn

9. - - Li UtA'L l*Uj V rIm rmi IUItLZ a) Degraded pewei r-%MA.M ,T%I fVi =~T AI ~ r il K. IULULPLtrrLut'~li~

a) Projected or mcasured activity at the sitc boundary, avcraged ovcr one hour, is greater than or equal to .05 r,__/hr TEDE or .15Cm>re hr Child Thyroeid CDE or liquid radiologic-al effluents exceed ODCM limits.

6-SAFF-SYEM FUNGN a) Dcviation from Tcchnical Specification action statemcnts when required to shutdown or roldown or deviatiens pursuant to 1OGFR50.51(x).

b) Loss of dose assessment capabilities.

c) Loss of comrmunic-ations.

Revision 29 D-12

TABLE D-2 (Gontinued)

U A7 A Q %C TI C-rA*Trfnh fllA-TfnKI n

-7. .... _ .. ^_ _ . _ _ .... _ _ .. _ . _. _ .. _ _ ..

a) Sccurity thrcat onsitc but outsidc thc Protceted Arca Sccurity Fcnce (c.g.

attempted cntry or sabotagc which has been stopped outsidc thc sccurity fenee).

b) Fire or cxplosion onsitc.

c) Aircaft crash, unusual aircraft activity, train derailmcnt, turbinc failurc, toxic er flammable gas.

8. NAThRAL EVENTS a) Tornado, flood, loss of Dardancilc Rcscrvoir, carthqual(c.

4. MISCELLANEOUS EVENTS a) Other plant conditions cxist that warrant increased awareness on the part of the eoprating staff and state and/or lecal effsit_autheritics or invelve ethce than a normal controlled shutdown.

Revision 29 D-13

TABLE D 2 (Ccntinued)

ALERT 4:-

nnfThAAflV I -- ---

I -

CVC-=R~A 1 w9

,1 v- .

MIE7KrrC

2. REACTOR COOLANT SYSTEM LEAKAGE a) RCS lcakagc > 11 gpm.

3. SECONDARY SYSTEM EVENTS a) S/C Tube Leak > 10 gpmn with an ongoing steam release A MI Cfr-MTIAt nn%~AIfEf CATI I MCC~

I-K7V..

M11*..M* I7'tVrLVrV M I7 a) Statien Blaekout.

b) Loss of all vital DG.

fl A rTfr% f~fT'Ag tI rs I gOr~r1-c' 5-. IS. *.~AJ.J1...

II I "W~.J.I V 42:

a) Projected or measured activity at the site beundary, averaged- oere one hour, is grcatcr than or equal to .5 rfrem/lhr TEDE or 1.5 mrem/lhr Child Thy-rneid GDE eF liqui Fadeleia e_^;fluent{ s -exceed^- 10ni s;_ rnrGM limits.

b) High radiation'airborne levels.

c CA E:=V I

CVC=RA uiu IN1rrTTnlI a) RPS failurc to cemplete an automatic trip.

b) Control Room evacuation.

C) LOes of decay heat remeval capabilities.

d) Loss of Control Room Annunciators-.

Revision 29 D-14

TABL6E D-2 (Gontinued)

LI A 7AfnrC -rf% -rAT-r~tNk f'Nntfl A-rTf~N a) Ongoing security thrcat within the Preteeted Arca Sccurity Fenee but outsidc of plant buildings.

b) Firc or explesien onsite affect~ing one train of ESF systeoms.

e) Aircr~aft cr-ash, missiles, toxic or flammable gas affecting onc train of ESF systems.

8. NATURAL EVENTS a) Tornado, high winids, flood, less efDardanelle Reseweir-, eaithguaI .

9. MISCELLANEOUS-EVENTS a) Gther plant conditions exist that warrant preeautionary activatien of the Tachniialnupport Gctcrg and placing thonearsitedpebatiens Emergency Fadit an _ ot her 1wey emr-gne _eerrl en o+tandhs Revision 29 D-15

TABLE D 2 (Continued)

~1TF rF mFr(prFNh-lc nrlTYAADV CVC r CKA tMIEr..r 47 a) Core damage indicated with an inadequate core cooling condition.

b) Gentainment radiation readings which indicatc LOCA and > 1% fuel claddin 2 ncREACTOR COOLANT SYSTEM LEAKAGE a) RnS ICakage > 14 gpm with ICC conditions.

3. SECOENDAR-Y SY-SM EVENTS a) S/G tubc rupture >41 gpm with an ongoing steam relcase and RCS Activity

4Lv Gigmnbut<38m ue~. ddieladdingf e).

ELERICAL 1. IA hFAILURES

,rER a) Blackoeut > 15 mninutes-.

a). Railgel F5l effluet aU;e grateVry tha 'F euall teUUI 5y mrmhrFD eFU;1-4f-'

b) rn~r~el Loss of ALL vital DG fer > 15 minutes.

Ghldl~ 4by-id GD atte ie endf-

5. RADIOLO)GIGAL EFFLUENTSc b) Spent fuel accident.

Revision 29 D-16

TABLE D-2 (Gentinuced) fi`'A E-rE'r%I 1'.ifer-ft A f-I I RIe#Tef~P.I h.

1. hi\ - t I Y bYh I tM PUNL I 1UN b) Control room cvacuation and control of shutdown systems net established in 15 minutes.

c) Loss of both S/Cs as a heat rcmoval method.

d) Loss of Control Room annunciators with a transient in progrcss.

7. HAZARDS TO STATION OPERATION a) Ongoing security thrcat within plant buildings but not within thc Control Roem or vital arcas.

b) Firc or cxplosion onsite affc-ting both trains of ESF Systems.

c) Aircraft crash, missiles, toxic or flammable gas affecting both redundant ESF trains.

V. NATURAL EVENTS a) Tornado, high winds, flood, loss of Dardancllc Rcscrvoir, carthquakc.

9. MISCELLANEOUS EVENTS a) Other plant conditions exist that warrant activation of thc cmergency rcsponse facilitics and monitoring tcams or a prccautionary notification to the publie near the sitc.

Revision 29 D-17

TABLE D 2 (Continued)

GENERAL EMERGENGY 1

. PPTMAPY ADV CVFM IrFFMC a)

-. f Containmcnt radiation rcadines . ___-- * . which

...... .. _.. ... indicatc _ __LOA

- .......... _ . -and >5O%-fuiel ovcrhcat.

b) Corc mclt with Containmcnt Integrit' Loet or Challenged.

c) Loes of or challengc to all three fission produ*t barricrs.

2. REACTOR COOLANT SYSTEM LEAKAGE

-N/

.SEGGNDARY-SYSTEM EVENw N/A

1. ELECTRICAL PO\WER FAILURES N/A

5. RADIOLOGICAL EFFLUENTS 6- SAFETY SSTEM FUNGTIQN N/A

7. HAZARDS TO STATION OPERATION a)

-I unoolna securit: thrcat within thc

- - - -- - - -d -- - - - -- -

_.. _Control

__. . _._.Room . .__. .or. _vital

. . . . areas.

Revision 29 D-18

TABLE D 2 (Continued)

I ATRIPAI EVEM N/A

9. MISCELLANEOUS EVENTS RPlant.

9 nditin exist that make release of large amount of radieaDtivity p-ssible.

Revision 29 D-19

Attachment 2 OCAN020407 Proposed Emergency Plan Changes - Changes Incorporated

TABLE D-1 Index of Emergency Action Levels M. .606------

M-l 15r41-1 I AG1 Offsite dose resulting from an actual or imminent release of gaseous radioactivity exceeds 1000 mR TEDE or 5000 mR child thyroid CDE for the actual or projected duration of the release using actual meteorology ASI Offsite dose resulting from an actual or imminent release of gaseous radioactivity exceeds 100 mR TEDE or 500 mR child thyroid CDE for the actual or projected duration of the release M1 Any UNPLANNED release of gaseous or liquid radioactivity to the environment exceeds 200 times the radiological effluent ODCM limits for 15 minutes or longer AA2 Damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel outside the reactor vessel AA3 Release of radioactive material or elevated radiation levels within the facility that impede operation of systems required to maintain safe operations or to establish or maintain cold shutdown AU1 Any UNPLANNED release of gaseous or liquid radioactivity to the environment that exceeds two times the radiological effluent ODCM limits for 60 minutes or longer AU2 Unexpected rise in plant radiation

!) e p l 06 X . ~~ i2"i-il'S 0 CG1 Loss of reactor vessel inventory affecting fuel clad integrity with containment challenged with irradiated fuel in the reactor vessel CS1 Loss of reactor vessel inventory affecting core decay heat removal capability CS2 Loss of reactor vessel inventory affecting core decay heat removal capability with irradiated fuel in the reactor vessel CA1 Loss of RCS inventory CA2 Loss of reactor vessel inventory with irradiated fuel in the reactor vessel CA3 Loss of all offsite power and loss of all onsite AC power to required 4.16KV busses CA4 nability to maintain plant in cold shutdown with irradiated fuel in the reactor vessel CU RCS leakage CU2 UNPLANNED loss of RCS inventory with irradiated fuel in the reactor vessel CU3 Loss of all offsite power to vital busses for greater than 15 minutes CU4 NPLANNED loss of decay heat removal capability with irradiated fuel in the reactor vessel CU5 Fuel clad degradation CU6 UNPLANNED loss of all onsite or offsite communications capabilities CU7 UNPLANNED loss of required DC power for greater than 15 minutes CU8 Inadvertent criticality Revision XX D-4

TABLE D-1 (Continued)

Index of Emergency Action Levels E-HU1 IDamage to a loaded cask CONFINEMENT BOUNDARY E-HU2 lConfirmed security event with potential loss of level of safety of the ISFSI

=- *6iam IM- =-a sI*

FG1 Loss of ANY two barriers AND loss or potential loss of third barrier FSI Loss or potential loss of ANY two barriers FAI ANY loss or ANY potential loss of EITHER fuel clad OR RCS FUl ANY loss or ANY potential loss of containment HG1 Security event resulting in loss of physical control of the facility HG2 Other conditions exist which in the judgment of the SM/TSC Director/EOF Director warrant declaration of General Emergency HS1 Confirmed security event in a plant VITAL AREA HS2 Other conditions exist which in the judgment of the SM/TSC Director/EOF Director warrant declaration of Site Area Emergency HS3 Control Room evacuation has been initiated and plant control cannot be established HAl Confirmed security event within a plant PROTECTED AREA HA2 Other conditions exist which in the judgment of the SM/TSC Director/EOF Director warrant declaration of an Alert HA3 Control Room evacuation has been initiated HA4 FIRE or EXPLOSION affecting the operability of plant safety systems required to establish or maintain safe shutdown HA5 Release of toxic or flammable gases within or adjacent to a VITAL AREA which jeopardizes operation of systems required to establish or maintain safe shutdown HA6 Natural and destructive phenomena affecting the plant VITAL AREA HUI Confirmed security event which indicates a potential degradation in the level of safety of the plant HU2 Other conditions exist which in the judgment of the SM/TSC Director/EOF Director warrant declaration of an NUE HU4 FIRE within PROTECTED AREA boundary not extinguished within 15 minutes of detection HU5 Release of toxic or flammable gases deemed detrimental to normal operation of the plant HU6 Natural and destructive phenomena affecting the PROTECTED AREA Revision XX D-5

TABLE D-1 (Continued)

Index of Emergency Action Levels A,- !zL .6 - - - -

SG1 Prolonged loss of all offsite power and prolonged loss of all onsite AC power to vital 4.16KV busses SG2 Failure of the Reactor Protection System to complete an automatic trip and manual trip was NOT successful and there is indication of an extreme challenge to the ability to cool the core SS1 Loss of all offsite power and loss of all onsite AC power to vital 4.16KV busses SS2 Failure of Reactor Protection System instrumentation to complete or initiate an automatic reactor trip once a Reactor Protection System setpoint has been exceeded and manual trip was NOT successful SS3 Loss of all vital DC power SS4 Complete loss of heat removal capability SS6 Inability to monitor a TRANSIENT in progress SA2 Failure of Reactor Protection System instrumentation to complete or initiate an automatic reactor trip once a Reactor Protection System setpoint has been exceeded and manual trip was successful SA4 UNPLANNED loss of most or all safety system annunciation or indication in control room with either (1) a PLANT TRANSIENT in progress, or (2) SPDS and PMS dynamic alarm functions are unavailable SA5 AC power capability to vital 4.16KV busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout SU Loss of all offsite power to vital 4.16KV busses for greater than 15 minutes SU2 Inability to reach required shutdown within Technical Specification limits SU3 UNPLANNED loss of most or all safety system annunciation or indication in the control room for greater than 15 minutes SU4 Fuel clad degradation SU5 RCS leakage SU6 UNPLANNED loss of all onsite or offsite communications capabilities SU8 Inadvertent criticality Revision XX D-6

Attachment 3 OCAN020407 Proposed EALs

PROC.MORK PLAN NO. PROCEDURENVORK PLAN TITLE: PAGE: 1 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-O SECTIONS PAGE NO.

1.0 Purpose........................................... .........

.................. 2 2.0 Scope............................................. .........

.................. 2 3.0 References........................................ .........

.................. 2 4.0 Definitions....................................... .........

.................. 3 5.0 Responsibility and Authority...................... . . . . . . . . . . . .. . . . . . ........

1 6.0 Instructions...................................... . . . . . . . . . . . . . . . . . . .........

1 ..........3 .

7.0 Attachments and Forms............................ .........

.................. 14 7.1 Attachment 1 - Index of EALs.............. .........

.................. 15 7.2 Attachment 2 - EAL Matrix................. .........

.................. 18 7.3 Attachment 3 - Emergency Action Levels

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 2 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 1.0 PURPOSE This procedure establishes criteria for detection and classification of plant events into the four standard Emergency Classes.

2.0 SCOPE This procedure is applicable to Units 1 and 2 in all modes; it does not include specific plant casualty procedures or systems operations requirements, but rather provides administrative processes only.

3.0 REFERENCES

3.1 REFERENCES

USED IN PROCEDURE PREPARATION:

3.1.1 ANO Emergency Plan 3.1.2 NUREG-0654/FEMA-REP-1, Rev. 1 3.1.3 10 CFR 50 3.1.4 NRC Branch Position on Acceptable Deviations to Appendix 1 to NUREG-0654/FEMA-REP-1, July 11, 1994 3.1.5 NEI 99-01 Rev. 4 Methodology for Development of Emergency Action Levels

3.2 REFERENCES

USED IN CONJUNCTION WITH THIS PROCEDURE:

3.2.1 LI-102, "Corrective Action Process" 3.2.2 1903.011, "Emergency Response/Notifications" 3.2.3 1903.064, "Emergency Response Facility - Control Room" 3.2.4 1903.065, "Emergency Response Facility - Technical Support Center (TSC)"

3.2.5 1903.066, "Emergency Response Facility - Operational Support Center (OSC)"

3.2.6 1903.067, "Emergency Response Facility - Emergency Operations Facility (EOF)"

3.2.7 1203.025, "Natural Emergencies" 3.2.8 2203.008, "Natural Emergencies" 3.2.9 1202.XXX, "Emergency Operating Procedures" 3.2.10 2202.XXX, "Emergency Operating Procedures" 3.2.11 1404.016, "Post Earthquake Data acquisition and Measurement"

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 3 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 3.2.12 1904.002, "Offsite Projections-RDACS Computer Method" 3.2.13 NRC Position Paper on "Timeliness of Classification of Emergency Conditions", dated August 17, 1995 3.2.14 1607.001, "Reactor Coolant System Sampling" 3.2.15 2607.001, "Unit 2 Reactor Coolant System Sampling" 3.3 RELATED ANO PROCEDURES:

3.3.1 1043.042, "Response to Contingencies" 3.3.2 1502.004, "Control of Unit 1 Refueling", Attachment H 3.3.3 1903.023, "Personnel Emergency" 3.3.4 ANO Security Plan/Security Procedures 3.3.5 1015.007, "Fire Brigade Organization and Responsibilities" 3.3.6 1903.042, "Duties of the Emergency Medical Team" 3.3.7 1903.043, "Duties of the Emergency Radiation Team" 3.4 REGULATORY CORRESPONDENCE CONTAINING NRC COMMITMENTS WHICH ARE IMPLEMENTED IN THIS PROCEDURE (DENOTED IN LEFT HAND MARGIN AND BY

[BOLD]:

3.4.1 OCAN068320 (P-10766) - Section 4.11 4.0 DEFINITIONS 4.1 Affecting Safe Shutdown: Event in progress has adversely affected functions that are necessary to bring the plant to and maintain it in the applicable HOT or COLD SHUTDOWN condition.

Plant condition applicability is determined by Technical Specification LCOs in effect.

Example 1: An event causes damage that results in entry into an LCO that requires the plant to be placed in HOT SHUTDOWN. HOT SHUTDOWN is achievable, but COLD SHUTDOWN is not. This event is not "AFFECTING SAFE SHUTDOWN."

Example 2: An event causes damage that results in entry into an LCO that requires the plant to be placed in COLD SHUTDOWN. HOTSHUTDOWN is achievable, but COLD SHUTDOWN is not. This event is "AFFECTING SAFE SHUTDOWN."

PROCJWORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 4 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-O 4.2 Bomb: refers to an explosive device suspected of having sufficient.force to damage plant systems or structures.

4.3 Civil Disturbance: A group of persons violently protesting station operations or activities at the site.

4.4 Confinement Boundary: The barrier(s) between areas containing radioactive substances and the environment.

4.5 Containment Closure (Unit 1): A condition where at least one integral barrier to the release of radioactive material is provided.

4.6 Containment Closure (Unit 2): Containment Closure is a preliminary action that immediately and effectively reduces the likelihood of a release while providing flexibility to-have the Containment Building open under appropriate conditions. The Containment Building provides the last integral barrier to the release-of radioactive material to the general public. During core alterations with-less than 23 feet of coolant above the fuel, Containment Closure must be set; with greater than 23 feet of coolant above the fuel, Containment Closure must be capable of being set. Containment Closure is set when the following conditions have been met:

The equipment hatch door is closed and held in place by a minimum of four bolts such that no gaps exist in the sealing surface or the Temporary Equipment Hatch Cover (TEHC) is installed per Temporary Equipment Hatch Cover Installation/Removal (2504.036).

A minimum of one barrier in each airlock is closed.

Each penetration providing access from Containment to the outside atmosphere, is closed by a valve, blank flange, or other approved closure mechanism. Opening of systems inside.

Containment may create a Containment breach potential that is NOT readily apparent. An example would be an opening of the S/G secondary side manways thus expanding closure concerns to piping and valves up to the MSIVs.

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 5of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 4.7 Containment Integrity (Unit 1): CONTAINMENT INTEGRITY shall exist when:

All penetrations required to be closed during accident conditions are either:

a. Capable of being closed by an OPERABLE containment automatic isolation valve system, or

b. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by Technical Specification 3.6.3.1.

All equipment hatches are closed and sealed.

Each airlock is OPERABLE pursuant to Technical Specification 3.6.2.

The containment leakage rates are within the limits of Technical Specification 3.6.1.

The sealing mechanism associated with each penetration (e.g.,

welds, bellows or O-rings) is OPERABLE.

4.8 Containment Integrity (Unit 2): CONTAINMENT INTEGRITY shall exist when:

All penetrations required.to be closed during accident conditions are either:

a. Capable of being closed by an OPERABLE containment automatic isolation valve system, or

b. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by Technical Specification 3.6.3.1.

All equipment hatches are closed and sealed.

Each airlock is OPERABLE pursuant to Technical Specification 3.6.1.3.

The containment leakage rates are within the limits of Technical Specification 3.6.1.2.

The sealing mechanism associated with each penetration (e.g.,

welds, bellows or O-rings) is OPERABLE.

4.9 Courtesy Call: A notification to the Arkansas Department of Health and follow-up notification to the NRC for conditions/events other than those constituting an Emergency Class as listed in procedure 1903.11, "Emergency Response/Notifications", Section 6.3.

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 6 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX.O 4.10 Emergency Action Level: Alarms, instrument readings or visual sightings that have exceeded pre-determined limits which would categorize the situation into-an initiating condition of one of the following four Emergency Classes:

Notification of Unusual Event (NUE)

Alert

Site Area Emergency (SAE)

General Emergency (GE) 4.10.1 Notification of Unusual Event: Unusual events are in progress or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

4.10.2 Alert: Events are in progress or have occurred which involve an actual or potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

4.10.3 Site Area Emergency: Events are in progress or have occurred which involve actual or likely major failures of plant functions needed for protection of the public.

Any releases are not expected to exceed EPA Protective Action Guideline exposure levels except near the site boundary.

4.10.4 General Emergency: Events are in progress or have occurred which involve actual or imminent substantial core degradation or melting with the potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels off'site for more than the immediate site area.

{0CAN068320) 4.11 (Emergency Direction and Control: Overall direction of facility response which must include the non-delegable responsibilities for the decision to notify and to recommend protective actions to Arkansas Department of Health personnel and other authorities responsible for offsite emergency measures. With activation of the EOF, the EOF Director typically assumes the responsibility for Emergency Direction and Control. The management of on-site facility activities to mitigate accident consequences remains with the TSC Director in the Technical Support Center. The Shift Manager retains responsibility for the Control Room and plant systems operation.]

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 7of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 4.12 Emergency Operations Facility (EOF): A near-site emergency response facility located approximately 0.65 miles northeast of the reactor buildings (the ANO Training Center).

4.13 Emergency Planning Zone (EPZ): The EPZ considered by this procedure is the inhalation zone and is that area within approximately a 10 mile radius of ANO.

4.14 Emergency Response Organization (ERO): The organization which is composed of the Initial Response Staff (IRS), the EOF staff, the TSC staff, the OSC staff, and the Emergency Team members. It has the capability to provide manpower and other resources necessary for immediate and long-term response to an emergency situation.

4.15 EPA Protective Action Guideline (PAG) Exposure Levels: The projected dose to reference man, or other defined individual, from an unplanned release of radioactive material at which a specific protective action to reduce or avoid that dose is recommended (i.e., 1 Rem TEDE or 5 Rem Child Thyroid (CDE)).

4.16 Exclusion Area: That area surrounding ANO within a minimum radius of 0.65 miles of the reactor buildings, but outside the protected area and controlled to the extent necessary by ANO during periods of emergency.

4.17 Explosion: A rapid, violent, unconfined combustion, or catastrophic failure of pressurized equipment that imparts energy of sufficient force to potentially damage permanent structures, systems, or components.

4.18 Extortion: An attempt to cause an action at the station by threat of force.

4.19 Faulted: In a steam generator, the existence of secondary side leakage that results in an uncontrolled decrease in steam generator pressure or the steam generator being completely depressurized.

4.20 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.

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

4.22 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.

PROCJWORK PLAN NO. PROCEDUREJWORK PLAN TITLE: PAGE: 8 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX.XX-0 4.23 Immediately Dangerous To Life and Health (IDLH): A condition that either poses an immediate threat to life and health or an immediate threat of severe exposure to contaminants which are likely to have adverse delayed effects on health.

4.24 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.

4.25 Initial Response Staff (IRS): The emergency organization composed of plant personnel which must be able to respond to the site in accordance with Table B-1 of the Emergency Plan.

4.26 Initiating Condition (IC): One of a predetermined subset of nuclear power plant conditions where either the potential exists for a radiological emergency, or such an emergency has occurred.

4.27 Intrusion/Intruder: A person(s) present in a specified area without authorization. Discovery of a BOMB in a specified area is indication of INTRUSION into that area by a HOSTILE FORCE.

4.28 Lower Flammability Limit (LFL): The minimum concentration of combustible substance that is capable of propagating a flame through a homogenous mixture of the combustible and a gaseous oxidizer.

4.29 Normal Plant Operations: Activities at the plant site associated with routine testing, maintenance, or equipment operations, in accordance with normal operating or administrative procedures.

Entry into abnormal or emergency operating procedures, or deviation from normal security or radiological controls posture, is a departure-from NORMAL PLANT OPERATIONS.

4.30 Normal Makeup (MU) Capacity: Normal MU capacity is defined as the maximum expected water addition to the RCS through the MU line with the letdown line isolated. This amount will vary with RCS pressure.

4.31 Offsite: Those areas outside the Exclusion Area boundary.

4.32 Onsite: The area within the Exclusion Area boundary.

4.33 Operational Support Center (OSC): Emergency response center within the ANO Maintenance Facility where support is coordinated for the following functions:

Onsite Radiological Monitoring

Maintenance

Nuclear Chemistry

Emergency Medical Support

Fire Fighting Support The OSC also serves as the briefing area for repair and damage control teams and is located in the Maintenance Facility.

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 9 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 4.34 PLANT TRANSIENT:

4.34.1 Any unplanned reactor trip from criticality.

4.34.2 A planned reactor trip in which the expected post-trip response did not occur.

4.34.3 Any event resulting in an automatic ESAS (Unit 1) or ESF (Unit 2) actuation or any event requiring manual initiation of these systems where automatic initiation would likely have occurred.

4.34.4. Any turbine-generator power change in excess of 100 MWe in less than one (1) minute other than a momentary spike due to a grid disturbance or a manually initiated runback.

4.34.5 Any unplanned main turbine or main feedwater pump turbine trip which results in a significant plant transient (change in excess of 100 MWe).

4.35 Protected Area: The area encompassed by physical barriers (i.e.,

the security fence) and to which access is controlled.

4.36 RCS Leakage: RCS leakage is defined as a loss of RCS inventory due to a leak in the RCS or a supporting system that is not or cannot be isolated within 10 minutes.

4.37 Ruptured: In a steam generator, existence of primary-to-.

secondary leakage of a magnitude sufficient to require or cause a reactor trip and safety injection.

4.38 Sabotage: Deliberate damage, mis-alignment, or mis-operation of plant equipment with the intent to render the equipment inoperable. Equipment *found tampered with or damaged due to malicious mischief may NOT meet the definition of SABOTAGE until this determination is made by security supervision.

4.39 Significant Transient: An UNPLANNED event involving one or more of the following: (1).automatic turbine runback greater than 25%

thermal reactor power, (2) electrical load rejection greater than 25% full electrical load, (3) Reactor Trip, (4) Safety Injection Activation, or (5) thermal power oscillations greater than 10%

4.40 Strike Action: A work stoppage within the PROTECTED AREA by a body of workers to enforce compliance with demands made on the company. The STRIKE ACTION must threaten to interrupt NORMAL PLANT OPERATIONs.

4.41 Technical Support Center (TSC): The location within the ANO Administration Building equipped with instrumentation and communication systems and facilities useful in monitoring the course of an accident; this center is located in the 3rd Floor of the ANO Administration Building.

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 10 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 4.42 Unplanned: A parameter change or an event that is not the result of an intended evolution and requires corrective or mitigative actions.

4.43 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.

4.44 Visible Damage: Damage to equipment or structure that is readily observable without measurements, testing, or analysis. Damage is sufficient to cause concern regarding the continued operability or reliability of affected safety structure, system, or component. Example damage includes: deformation due to heat or impact, denting, penetration, rupture, cracking, paint blistering. Surface blemishes (e.g., paint chipping, scratches) should not be included.

PROC.MORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 11 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 4.45 Vital Area: Any area, normally within the PROTECTED AREA, which contains equipment, systems, components, or material, the failure, destruction, or release of which could directly or indirectly endanger the public health and safety by exposure to radiation.

5.0 RESPONSIBILITY AND AUTHORITY 5.1 The responsibility for event classification is assigned to the individual with responsibility for Emergency Direction and Control (i.e., The Shift Manager, TSC Director, or EOF Director).

5.2 The Control Room Supervisor (CRS) will assume Emergency Direction and Control responsibilities if the SM is not available to assume this responsibility (e.g. the SM becomes incapacitated and a replacement has not yet arrived).

5.3 Any individual who observes an initiating condition which warrants an emergency class declaration, as described in Attachment 3, shall immediately notify the person with current responsibility for Emergency Direction and Control (i.e. SM/TSC Director/EOF Director).

6.0 INSTRUCTIONS NOTE On emergencies that affect both units such as earthquakes, tornadoes, etc., the unit with the highest Emergency Classification should declare the emergency.

6.1 CLASSIFYING EMERGENCIES:

1 NOTE NRC guidelines recommend that once indications are available to ANO staff that

.an EAL has been exceeded, a 15 minute goal is a reasonable period of time for assessing and classifying an emergency.

6.1.1 When indications of abnormal occurrences are received by the Control Room staff, the SM/TSC Director/EOF Director shall:

A. Verify the indications of the off-normal event or reported sighting.

B. Ensure that the immediate actions (e.g., use of Emergency and Abnormal Operating Procedures) are taken for the safe and proper operation of the plant.

C. Compare the abnormal conditions with those listed in the "Index of Emergency Action Levels".

D. Turn to the appropriate tab which corresponds to the condition picked from the Index of EALs.

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 12 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 .

E. Assess the information available from valid indications or reports, then:

1. Compare information to criteria given for EAL.

2. Review any Related EALs to determine if the abnormal conditions meet those criteria.

3. Declare the emergency classification that is indicated. If it appears that different classifications could be made for the current plant conditions, the highest classification indicated should be the one that is declared.

NOTE The emergency action levels described in this procedure are not intended to be used during maintenance and/or testing situations where abnormal temperature, pressure, equipment status, etc., is expected. In addition, each EAL contains nformation on the mode(s) of operation during which it is applicable.

6.1.2 Due to the speed in which events sometimes progress and the duty of the plant operators to take immediate corrective actions, an event may occur which was classifiable as an emergency, however, prior to offsite notifications the corrective actions taken may have removed the conditions that would have resulted in an emergency declaration. In this situation, it is not necessary to make an actual declaration of the emergency class, but an ENS notification to the NRC within one hour of the discovery of the undeclared event will provide an acceptable alternative. A courtesy call shall be made to ADH. Subsequent activation of response organization should be based upon the current plant conditions.

6.1.3 If no emergency declaration is required, then refer to procedure 1903.011, "Emergency Response/Notifications",

Section 6.4, to determine if the event warrants a "For Information Only" notification to Entergy Management, NRC Resident Inspector and/or the Arkansas Department of Health.

6.1.4 Upon declaration of an emergency classification implement procedure, 1903.011, "Emergency Response Notifications", to ensure that immediate notification requirements are met and the proper Emergency Plan response is taken.

6.1.5 Upgrade the emergency classification if plant conditions degrade per steps 6.1.1.A through E.

6.1.6 Downgrade the emergency classification when plant conditions have improved and step 6.2 is applicable.

PROC.MORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 13 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-.

6.2 DOWNGRADING THE EMERGENCY CLASSIFICATION:

6.2.1 Assess the current plant conditions, then perform the following:

A. Compare the abnormal conditions with those listed in the "Index of Emergency Action Levels".

B. Turn to the appropriate tab which corresponds to the condition picked from the Index of EALs.

C. Assess the information available from valid indications or reports; compare it to the given EALs. Obtain concurrence from NRC and State officials that downgrading is appropriate (if their emergency response organizations have been activated as a result of this event). Downgrade to the emergency classification that is indicated.

D. If the indications or reports do not match the given EALs, then refer to the Miscellaneous Tab and using appropriate judgment, determine if the plant status warrants downgrading the emergency classification.

6.2.2 Perform notifications to downgrade the emergency classification if appropriate per procedure 1903.011, "Emergency Response/Notifications".

6.2.3 If no emergency classification appears necessary, then terminate the emergency per step 6.3.

6.2.4 If the emergency classification is still required, repeat steps 6.2.1 through 6.2.3 whenever plant conditions again appear to have improved.

6.3 TERMINATING THE EMERGENCY:

6.3.1 Compare the existing plant conditions with the following:

A. Plant conditions no longer meet the emergency action level criteria AND it appears unlikely that current conditions will degrade further requiring reinstitution of an emergency classification.

B. Non-routine releases of radioactive material to the environment are under control or terminated.

C. Any fire, flood, earthquake, or similar emergency condition is controlled or has ceased.

PROC./WORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 14 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-O D. All specified corrective actions have occurred OR the plant has been placed in the appropriate operational mode.

E. All required notifications have been completed.

F. NRC and State officials are in agreement that termination or transition to the recovery phase is appropriate (if their emergency response organizations have been activated as a result of this event).

6.3.2 IF the conditions of 6.3.1 A-F are met, THEN terminate the emergency or proceed to the recovery phase.

7.0 ATTACHMENTS AND FORMS 7.1 Attachment 1: Index of EALs 7.2 Attachment 2: EAL Matrix 7.3 Attachment 3: Emergency Action Levels

PROC.JORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 15 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 ATTACHMENT 1 NOTE Once available plant parameters reach an Emergency Action Level (EAL),

classifications should be made within 15 minutes.

Index of Emergency Action Levels 4E 4 . .4 .0 6 . _ -

e -*- __ _ _. __ _ _ _. _ e -. -.........

* *4 *4 6 - M - 6. - B y 0 M ' Any UNPLANNED Release of Gaseous or. Liquid Radioactivity tothe Environment Pg. 42 ASI1 that Exceeds 200 Times the Radiological Effluent ODCM limits for 15 Minutes or

.. Lo ng e r...'::.:-:'--..';-.'-'::--'-:;::'.'.,-

AA2 Damage to irradiated fuel or loss of water level that has or will result in Pg..45 AS.

uncovering irradiated fuel.'

AM .- Release of Radioactive Material or elevated Radiation Levels Within the Facility g. 47,.

That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintaini Cold Shutdown AU1 Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment Pg. 37 AA1 that Exceeds Two Times the Radiological Effluent ODCM limits for 60 Minutes or Longer.

AU2 Unexpected rise in plant radiation Pg. 40 - 3 A-.4..- ,.=D CA1 Loss of RCS inventory. .- . Pg. 65 CS CA2 Loss of reactor vessel inventory with'irradiated fuel in the reactor vessel. Pg. 67 CS2 '

CA3- Loss of all offsite power and loss of all onsite AC power to required 4.16KV busses Pg. 69 _AS1 HS1 CA4 Inability to maintain plant in cold shutdown with irradiated fuel in the reactor vessel. Pg. 70 . CS' CUI RCS leakage. Pg. 55 CA1 CA3 CU2 UNPLANNED loss of RCS inventory with irradiated fuel in the reactor vessel. . Pg. 56 CA2 CA4 CU3 Loss of all offsite power to vital busses for greater than 15 minutes Pg. 58l CU4 UNPLANNED loss of decay heat removal capability with irradiated fuel in the Pg. 59 CA1 CA3 reactor vessel.

CU5 Fuel clad degradation Pg. 61 CU6 UNPLANNED loss of all onsite or offsite communications capabilities. Pg. 62l CU7 UNPLANNED loss of required DC power for greater than 15 minutes. Pg. 63 CA3 CU8 Inadvertent criticality. Pg. 64 HA2

PROC./WORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 16 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 ATTACHMENT 1 NOTE Once available plant parameters reach an Emergency Action Level (EAL),

classifications should be made within 15 minutes.

Index of Emergency Action Levels E-HU1 Damage to a loaded cask CONFINEMENT BOUNDARY. Pg. 79 E-HU2 Confirmed security event with potential loss of level of safety of the ISFSI Pg. 81 i 1

0AS - J . I . =: .

FA1 ANY loss or ANY potential loss of EITHER fuel clad OR RCS. Pg. 83 FUI lANY loss or ANY potential loss of containment. Pg.82 11 ws 0 - - *5 _1 M

- S Sn S; -e - -,fi ,S - - ! i~ Cat _A r.-

iFil HA1 Confirmed security event within a plant PROTECTED AREA. Pg. 116 HS1i' HA2 Other conditions existing 'which in the judgment of the SM/TSC Director/EOF Pg. 117 Directorwarrant declaration of an Alert.:-:

HA3 Control Room evacuation has'been initiated. its.1.. :: Pg. 118 l: HS3 HA4 FIRE or EXPLOSION affecting'the'operabilityof plant safety systems required to Pg. 119 SS1FS1 AS1 establish or maintain safe shutdown.-: HS2 HA5. Release of toxic or flammable gases within or adjacent to a VITAL AREA which: -: Pg. 121 SS1 FS1 AS1 Jeopardizes operation of systems required to establish or maintain safe shutdown. HS2 HA6' Natural and destructive phenomena-affecting the plant VITAL AREA.,. Pg. 123 HU1 Confirmed security event which indicates a potential degradation in the level of Pg. 110 HA1 safety of the plant.

HU2 Other conditions existing which in the judgment of the SMITSC Director/EOF Pg. 111 Director warrant declaration of an NUE.

HU4 FIRE within PROTECTED AREA boundary not extinguished within 15 Minutes of Pg. 112 HA4 detection.

HU5 Release of toxic or flammable gases deemed detrimental to normal operation of Pg. 113 HA3 the plant. ' -_l_l HU6 INatural and destructive phenomena affecting the PROTECTED AREA. Pg. 114 HA6

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 17 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 ATTACHMENT 1 NOTE Once available plant parameters reach an Emergency Action Level (EAL),

classifications should be made within 15 minutes.

Index of Emergency Action Levels 0 0 .. ' at.I.e..

h - 0 0

0 - - 0 -- 1 1 - _ i - -

.~~~~ "" ! - ..

SA2 Failure of Reactor Protection System instrumentation to complete or initiate an : Pg. 141 SS2 automatic reactor trip once a reactor protection system setpoint has been -- ...

exceeded and manual trip was successful. :-_I.l -l--:-

SA4 UNPLANNED loss of 'Most or all safety system annunciation or indication in control Pg. 142 SS6 room with either (1) a PLANT TRANSIENT in progress, or (2) SPDS and PMS dynamic alarm functions are unavailable.

SA5 AC powercapabilitytoVital4.16KVbussesreducedtoasinglepowersourcefor Pg.- 144 l 51 greater than 15 minutes such that any additional single failure would result in station blackout. -

SU1 Loss of all offsite power to Vital 4.16KV busses for greater than 15 Minutes. Pg. 131 SA5 SU2 Inability to reach required shutdown within Technical Specification limits. PG. 133 SU3 UNPLANNED loss of most or all safety system annunciation or indication in the Pg. 134 SA4 control room for greater than 15 minutes.

SU4 Fuel clad degradation. Pg. 136 SU5 RCS leakage. Pg. 137 FA1 SU6 UNPLANNED loss of all onsite or offsite communications capabilities. Pg. 138 SU8 Inadvertent criticality. Pg. 140 FA1 HA2

PROC.IWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 37 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-O Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AU1 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Any UNPLANNED release of gaseous or liquid radioactivity to the environment that exceeds two times the radiological effluent ODCM limits for 60 minutes or longer Operating Mode Applicability:

All Emergency Action Level: 1 OR 2 OR 3 OR 4 Note: If monitor reading is sustained for the time period indicated in the EAL AND the required assessments using procedure calculations cannot be completed within this period, declaration must be made based on the valid radiation monitor reading.

1. VALID reading on any effluent monitor that exceeds two times the alarm setpoint established by a current release permit for 60 minutes or longer.

EFFLUENT MONITORS - UNIT 1 RX-9820 Containment Purge (channel 7 or 9)

RE-4830 Waste Gas Radiation Monitor RE-4642 Liquid Radwaste Monitor EFFLUENT MONITORS - UNIT 2 2RX-9820 Containment Purge (Channel 7 or 9) 2RE-2429 Waste Gas Monitoring System 2RE-2330 BMS Liquid Discharge Monitor 2RE-4423 Radwaste Liquid Discharge Monitor

2. VALID reading on Channel 7 of one or more of the following radiation monitors that exceeds the reading shown for 60 minutes or longer during a discharge:

MONITORS - Unit 1 LIMIT RX-9820 Containment Purge 5.90E-2 (pCi/cc)

RX-9825 Radwaste Area 5.36E-2 (,uCi/cc)

RX-9830 Fuel Handling Area 4.54E-2 (iiCi/cc)

RX-9835 Emergency Penetration Room 9.56E-1 (OCi/cc)

MONITORS - Unit 2 LIMIT 2RX-9820 Containment Purge 4.46E-2 (pCi/cc) 2RX-9825 Radwaste Area 3.32E-2 (pCi/cc) 2RX-9830 Fuel Handling Area 4.46E-2 (pCi/cc) 2RX-9835 Emergency Penetration Room 8.84E-1 (pCi/cc) 2RX-9840 Post Accident Sampling Building 4.42E-1 (pCi/cc) 2RX-9845 Aux. Building Extension 1.26E-1 (pCi/cc) 2RX-9850 Low Level Radwaste Storage Building 1.77E-1 (pCi/cc)

PROCJWORK PLAN N

O. PROCEDURE

NVORK PLAN TITLE: PAGE: 38 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT (AU1)

3. Confirmed grab sample analyses for gaseous or liquid releases indicates concentrations or release rates, with a release duration of longer than 60 minutes, in excess of two times the applicable values of the ODCM.

4. RDACS data indicating NUE.

Basis:

This IC addresses a potential or actual drop in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. ANO incorporates features intended to control the release of radioactive effluents to the environment.

Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. These controls are located in the Offsite Dose Calculation Manual (ODCM). The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of degradation in these features and/or controls.

The ODCM multiples are specified in AU1 and AA1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, NOT the magnitude of the associated dose or dose rate.

The ODCM contains the site specific release limits and appropriate surveillance requirements which normally monitor these limits. Releases should not be prorated or averaged over 60 minutes. For example, a release exceeding 4 times ODCM limits for 30 minutes does not meet the threshold for this IC. The one hour time period allows sufficient time to isolate any release after exceeding ODCM limits. Release continuing for more than one hour represents inability to isolate or control the release. The SM/TSC Director/EOF Director should not wait until 60 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 60 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the SM/TSC Director/EOF Director should, in the absence of data to the contrary, assume that the release has exceeded 60 minutes.

"UNPLANNED", as used in this context, includes any release for which a liquid waste release or a gaseous waste release discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm set points, etc.) on the applicable package permit. Unplanned releases in excess of two times of the ODCM limit that continue for 60 minutes or longer represent an uncontrolled situation and a potential degradation in the level of safety. It is not intended that the release be averaged over 60 minutes. The event should be declared as soon as it is determined that the release duration has or will likely exceed one hour.

PROC.iWORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 39 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT (AU1)

EAL #1 addresses radioactivity releases, that for whatever reason, cause effluent radiation monitor readings to exceed two times the alarm setpoint and releases are not terminated within 60 minutes. This alarm setpoint may be associated with a planned batch release, or a continuous release path. In either case, the setpoint is established by the discharge permit to warn of a release that is not in compliance.

EAL #2 is similar to EAL #1, but is intended to address effluent or accident radiation monitors on release pathways for which a discharge permit would not be prepared for a non-routine release. The ODCM establishes a methodology for determining effluent radiation monitor setpoints. The ODCM specifies default source terms from SAR and, for gaseous releases, prescribes the use of pre-determined annual average meteorology in the most limiting downwind sector for showing compliance with the regulatory commitments. These monitor reading EALs have been determined using this methodology.

EAL #3 addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, (e.g., spills of radioactive liquids into storm drains, leakage into the river water systems or lake,. etc.).

EAL #4 addresses RDACS calculation for NUE. RDACS is a 60 minute rolling calculation and once alarmed no additional 60 minutes are required.

Escalation is via AA1, AS1, or AG1.

PROCJWORK PLAN NO. PROCEDUREWIORK PLAN TITLE: PAGE: 40 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AU2 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Unexpected rise in plant radiation Operating Mode Applicability:

All Emergency Action Level: 1 OR 2

1. VALID indication of uncontrolled water level drop in the refueling canal or.

spent fuel pool with all irradiated fuel assemblies remaining covered by water.

AN Unplanned VALID Area Radiation Monitor reading rise Unit 1 RE-8009 Spent Fuel Area RE-8017 Fuel Handling Area Unit 2 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8912 Containment Incore Instrumentation

2. Unplanned VALID Area Radiation Monitor readings rise by a factor of 1000 over normal* levels.

Normal levels can be considered as the highest reading in the past twenty-four hours excluding the current peak value.

Basis:

All of the above events tend to have long lead times relative to a potential for radiological release outside the site boundary; thus impact to public health and safety is very low.

This IC addresses elevated radiation levels as a result of lowered water level above the reactor vessel flange or events that have resulted, or may result, in unexpected rises in radiation dose rates within plant buildings. These radiation rises represent a loss of control over radioactive material and may represent a potential degradation in the level of safety of the plant.

PROC.IWORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 41 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT (AU2)

In light of reactor cavity seal failure incidents, explicit coverage of these types of events via EAL #1 is appropriate given their potential for higher doses to plant staff. Specific indications may include instrumentation such as water level and local area radiation monitors, and personnel (e.g., refueling crew) reports. Depending on available level instrumentation, the declaration threshold may be based on indications of water makeup rate or drop in BWST (Unit 1) or RWT (Unit 2) level.

Classification as an Unusual Event is warranted as a precursor to a more serious event.

While a radiation monitor could detect a rise in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered. For example, the reading on an area radiation monitor located on the refueling bridge may rise due to planned evolutions such as head lift, or even a fuel assembly being raised in the manipulator mast. Generally, higher radiation monitor indications will need to be combined with another indicator (or personnel report) of water loss. For refueling events where the water level drops below the reactor vessel flange, classification would be via CU2.

This event escalates to an Alert per AA2 if irradiated fuel outside the reactor vessel is uncovered. For events involving irradiated fuel in the reactor vessel, escalation would be via the Fission Product Barrier matrix for events in operating modes 1-4.

EAL #2 addresses UNPLANNED rises in in-plant radiation levels that represent degradation in the control of radioactive material, and representa potential degradation in the level of safety of the plant. Normal levels can be considered as the highest reading in the past twenty-four hours excluding the current peak value.

This event escalates-to an Alert per AA3 if the rise in dose rates impedes personnel access necessary for safe operation.

PROC.JORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 42 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION C CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AA1 ALERT Initiating Condition:

Any UNPLANNED release of gaseous or liquid radioactivity to the environment exceeds 200 times the radiological effluent ODCM limits for 15 minutes or longer Operating Mode Applicability:

All Emergency Action Level: 1 OR 2 OR 3 OR 4 Note: If monitor reading is sustained for the time period indicated in the EAL AND the required assessments using procedure calculations cannot be completed within this period, declaration must be made based on the valid radiation monitor reading.

1. VALID reading on any effluent monitor that exceeds 200 times the alarm setpoint established by a current release permit for 15 minutes or longer.

EFFLUENT MONITORS - UNIT 1 RX-9820 Containment Purge (Channel 7 or 9)

RE-4830 Waste Gas Radiation Monitor RE-4642 Liquid Radwaste Monitor EFFLUENT MONITORS - UNIT 2 2RX-9820 Containment Purge (Channel 7 or 9) 2RE-2429 Waste Gas Monitoring System 2RE-2330 BMS Liquid Discharge Monitor 2RE-4423 Radwaste Liquid Discharge Monitor

2. VALID reading on Channel 7 of one or more of the following radiation monitors that exceeds the reading shown for 15 minutes or longer:

MONITORS - Unit 1 LIMIT RX-9820 Containment Purge 5.90E0 (pCi/cc)

RX-9830 Fuel Handling Area 4.54E0 (,uCi/cc)

RX-9825 Radwaste Area 5.36E0 (pCi/cc)

RX-9835 Emergency Penetration Room 9.56E+1 (pCi/cc)

MONITORS - Unit 2 LIMIT 2RX-9820 Containment Purge 4.46E0 (lICi/cc) 2RX-9825 Radwaste Area 3.32E0 (liCi/cc) 2RX-9830 Fuel Handling Area 4.46E0 (pCi/cc) 2RX-9835 Emergency Penetration Room 8.84E+1 (pCi/cc) 2RX-9840 Post Accident Sampling Building 4.42E+1 (pCi/cc) 2RX-9845 Aux. Building Extension 1.26E+l (pCi/cc) 2RX-9850 Low Level Radwaste Storage Building 1.77E+l (pCi/cc)

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JWORK PLAN TITLE: PAGE: 43 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT (AA1)

3. Confirmed grab sample analyses for gaseous or liquid releases indicates concentrations or release rates, with a release duration of 15 minutes or longer, in excess of 200 times the applicable values of the ODCM.

4. RDACS data indicating ALERT.

Basis:

This event escalates from the Unusual Event by escalating the magnitude of the release.

These EALs address a potential or actual drop in the level of safety of the plant as indicated by a radiological release that exceeds regulatory commitments for an extended period of time. ANO incorporates features intended to control the release of radioactive effluents to the environment. Further, there are administrative controls established to prevent unintentional releases, or control and monitor intentional releases. These controls are located in the ODCM. The occurrence of extended, uncontrolled radioactive releases to the environment is indicative of degradation in these features and/or controls.

The ODCM multiples are specified in AA1 and AU1 only to distinguish between non-emergency conditions, and from each other. While these multiples obviously correspond to an offsite dose or dose rate, the emphasis in classifying these events is the degradation in the level of safety of the plant, NOT the magnitude of the associated dose or dose rate.

Releases should not be prorated or averaged. For example, a release exceeding 400 times ODCM limits for 7.5 minutes does not meet the threshold for this event classification.

"UNPLANNED", as used in this context, includes any release for which a liquid waste release or a gaseous waste release discharge permit was not prepared, or a release that exceeds the conditions (e.g., minimum dilution flow, maximum discharge flow, alarm set points, etc.) on the applicable package permit. The SM/TSC Director/EOF Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 15 minutes. Also, if an ongoing release is detected and the starting time for that release is unknown, the SM/TSC Director/EOF Director should, in the absence of data to the contrary, assume that the release has exceeded 15 minutes.

EAL #1 addresses radioactivity releases that, for whatever reason, cause effluent radiation monitor readings to exceed 200 times the alarm setpoint and are not terminated within 15 minutes. This alarm setpoint may be associated with a planned batch release, or a continuous release path. In either case, the setpoint is established by the discharge permit to warn of a release that is not in compliance.

PROCJWORK PLAN NO. PROCEDURENJORK PLAN TITLE: PAGE: 44 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT (AA1)

EAL #2 is similar to EAL #1, but is intended to address effluent or accident radiation monitors on release pathways for which a discharge permit would not be prepared for a non-routine release. The ODCM establishes a methodology for determining effluent radiation monitor setpoints. The ODCM specifies default source terms from SAR and, for gaseous releases, prescribes the use of predetermined annual average meteorology in the most limiting downwind sector for showing compliance with the regulatory commitments. These monitor reading EALs have been determined using this methodology.

EAL #3 addresses uncontrolled releases that are detected by sample analyses, particularly on unmonitored pathways, e.g., spills of radioactive liquids into storm drains, leakage into Lake Dardanelle, etc.

EALs #1 and #2 directly correlate with the ODCM since annual average meteorology is required to be used in showing compliance with the ODCM and is used in calculating the alarm setpoints. The fundamental basis of these ICs is not a dose or dose rate, but rather the degradation in the level of safety of the plant implied by the uncontrolled release that was not isolated within 15 minutes.

Due to the uncertainty associated with meteorology, emergency implementing procedures should call for the timely performance of dose assessments using actual (real-time'and sector) meteorology in the event of a gaseous radioactivity release of this magnitude. The results of these assessments should be compared to AS1 and AGi to determine if the event classification should be escalated. Classification should not be delayed pending the results of these dose assessments.

EAL #4 addresses RDACS calculations for ALERT. Once RDACS data indicates ALERT, no additional time is required.

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 45 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AA2 ALERT Initiating Condition:

Damage to irradiated fuel or loss of water level that has or will result in the uncovering of irradiated fuel outside the reactor vessel Operating Mode Applicability:

All Emergency Action Level: 1 OR 2

1. A VALID alarm on one or more of the following radiation monitors:

Unit 1 RX-9820 Containment Purge (Channel 7 or 9)

RX-9825 Radwaste Area (Channel 7 or 9)

RX-9830 Fuel Handling Area (Channel 7 or 9)

RE-8060 Containment High Range Radiation Monitors RE-8061 Containment High Range Radiation Monitors RE-8009 Spent Fuel Area' RE-8017 Fuel Handling.

Unit 2 2RX-9820 Containment Purge (Channel 7 or 9) 2RX-9825 Radwaste Area (Channel 7 or 9) 2RX-9830 Fuel Handling Area (Channel 7 or 9) 2RE-8925-1 Containment High Range Radiation Monitors 2RE-8925-2 Containment High Range Radiation Monitors 2RE-8914 Spent Fuel Area 2RE-8915 Spent Fuel Area 2RE-8916 Spent Fuel Area 2RE-8912 Containment Incore Inst.

2. Water level drop in the refueling canal or spent fuel pool exceeds makeup capacity such that irradiated fuel has or will become uncovered.

Basis:

This IC, and associated EALs, address specific events that have resulted, or may result in unexpected rises in radiation dose rates within plant buildings, and may be a precursor to a radioactivity release to the environment. These events represent a loss of control over radioactive material and represent a degradation in the level of safety of the plant. These events escalate from AU2 in that fuel activity has been released, or is anticipated due to fuel heatup.

PROC.IWORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 46 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT (AA2)

These EALs apply to spent fuel requiring water coverage. There is time available to take corrective actions, and there is little potential for substantial fuel damage. Uncontrolled lowering of water level may be detected by visual observation, elevated radiation levels, or various other symptoms that consider valid indicators of the event. Fuel uncovery may be expected based on abnormal radiation level, visual observation, or best judgment of the SM/TSC Director/EOF Director based on present and past trends.

EAL #1 addresses radiation monitor indications of fuel uncovery and/or fuel damage. Elevated readings on ventilation monitors may be indicative of a radioactivity release from the fuel, confirming that damage has occurred.

Elevated background at the monitor due to water level drop may mask elevated ventilation exhaust airborne activity and should be considered. While a radiation monitor could detect a rise in dose rate due to a drop in the water level, it might not be a reliable indication of whether or not the fuel is covered. For example, the monitor could in fact be properly responding to a known event involving transfer or relocation of a source stored in or near the fuel pool or responding to a planned evolution such as removal of the reactor head. Application of these ICs requires understanding of the actual radiological conditions present in the vicinity of the monitor.

EAL #2 indicators may include instrumentation (such as water level and local area radiation monitors) and personnel (e.g., refueling crew) reports.

Depending on available level indication, the declaration threshold may need to be based on indications of water makeup rate or lowering in BWST (Unit 1) or RWT (Unit 2) level.

Escalation, if appropriate, would occur via AS1 or AG. or SM/TSC Director/EOF Director judgment.

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MWORK PLAN TITLE: PAGE: 47 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX.XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AA3 ALERT Initiating Condition:

Release of radioactive material or elevated radiation levels within the facility that impede operation of systems required to maintain safe operations or to establish or maintain cold shutdown Operating Mode Applicability:

All Emergency Action Level: 1 OR 2

1. VALID radiation readings greater than 15 mR/hr in areas requiring continuous occupancy to maintain plant safety functions such as Control Room/TSC, Controlled Access Area entry control point, Security Central Alarm Station (CAS).

2. VALID radiation readings greater than 5000 mR/hr in plant vital areas requiring infrequent access to maintain plant safety functions and'access is required for safe plant operation, but is impeded due to radiation dose rates.

Basis:

This IC addresses elevated radiation levels that impede necessary access to operating stations, or other areas containing equipment that must be operated manually or that requires local monitoring, in order to maintain safe operation or perform a safe shutdown. It is this impaired ability to operate the plant that results in the actual or potential substantial degradation of the level of safety of the plant. The cause and/or magnitude of the rise in radiation levels is not a concern of these EALs. The SM/TSC Director/EOF Director must consider the source or cause of the elevated radiation levels and determine if any other EAL may be involved. For example, a 15 mR/hr dose rate in the control room or a high radiation monitor reading may be a problem in itself.

However, the elevated radiation readings levels may also be indicative of high dose rates in the containment due to a LOCA. In this latter case, an SAE or GE may be indicated by the fission product barrier matrix EALs.

This IC is not meant to apply to elevated radiation levels in the containment as these are events which are addressed in the fission product barrier matrix EALs. This IC is not intended to apply to anticipated temporary rises due to planned events (e.g., incore detector movement, radwaste container movement, depleted resin transfers, etc.).

PROC.MORK PLAN NO. PROCEDURE/NORK PLAN TITLE: PAGE: 48 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT (AA3)

Areas requiring continuous occupancy include the control room and any other control stations that are manned continuously, such as CA-1 or CAS. The value of l5mR/hr is derived from the GDC 19 value of 5 Rem in 30 days with adjustment for expected occupancy times. Although Section III.D.3 of NUREG-0737, "Clarification of TMI Action Plan Requirements", provides that the 15 mR/hr value can be averaged over the 30 days, the value is used here without averaging, as a 30 day duration implies an event potentially more significant than an Alert.

For areas requiring infrequent access, the single value of 5000 mR/hr was selected because it is based on radiation levels which result in exposure control measures intended to maintain doses within normal occupational exposure guidelines and limits (i.e., 10 CFR 20), and in doing so, will impede necessary access. As used here, "impede" includes hindering or interfering provided that the interference or delay is sufficient to significantly threaten the safe operation of the plant. Stay times for levels up to that value are generally several minutes; enough time to enter an area and manually operate the equipment in order to maintain safe operation or perform a safe shutdown. The magnitude of the rise in radiation levels is not a concern of these EALs. The SM/TSC Director/EOF Director must consider the source or cause of the elevated radiation levels and determine if any other EAL may be involved.

Applicable areas requiring infrequent access are identified in the site's Abnormal Operating Procedures, Emergency Operating Procedures, the 10 CFR 50 Appendix R analysis, and/or the analyses performed in response to Section 2.1.6b of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-term Recommendations", when identifying areas containing safe shutdown equipment.

PROCJWORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 49 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT ASI SITE AREA EMERGENCY Initiating Condition:

Offsite dose resulting from an actual or imminent release of gaseous.

radioactivity exceeds 100 mR TEDE or 500 mR child thyroid CDE for the actual or projected duration of the release Operating Mode Applicability:

All Emergency Action Level: 1 OR 2 OR 3 Note: If dose assessment results are available at the time of declaration, the classification should-be based on EAL #2 instead of EAL #1. While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated/completed in order to determine if the classification should be subsequently escalated.

1. VALID reading on Channel 9 of one or more of the following radiation monitors that exceeds or is expected to exceed the reading shown for 15 minutes or longer:

MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E+l (lCi/cc)

RX-9830 Fuel Handling Area 4.54E+l (pCi/cc)

RX-9825 Radwaste Area 5.36E+l (1Ci/cc)

RX-9835 Emergency Penetration-Room 9.56E+2 (1.Ci/cc)

MONITORS - UNIT 2 LIMIT 2RX-9820 Containment Purge 4.46E+1 (pCi/cc) 2RX-9825 Radwaste Area 3.32E+1 (pCi/cc) 2RX-9830 Fuel Handling Area 4.46E+1 (pCi/cc) 2RX-9835 Emergency Penetration Room 8.84E+2 (pCi/cc) 2RX-9840 Post Accident Sampling Building 4.42E+2 (pCi/cc) 2RX-9845 Aux. Building Extension 1.26E+2 (pCi/cc) 2RX-9850 Low Level Radwaste Storage Building 1.77E+2 (pCi/cc)

2. Dose assessment using actual meteorology indicates doses greater than 100 mR TEDE or 500 mR child thyroid CDE at or beyond the site boundary.

3. Field survey results indicate closed-window dose rates exceeding 100 mR/hr expected to continue for more than one hour; or analyses of field survey samples indicate child thyroid CDE of 500 mR for 60 minutes of inhalation, at or beyond the site boundary.

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 50 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0O Attachment 3-ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT (ASI)

Basis:

This IC addresses radioactivity releases that result in doses at or beyond the site boundary that exceed a small fraction of the EPA Protective Action Guides (PAGs). Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public. While these failures may be addressed by other ICs, this IC provides appropriate diversity and addresses events which may not be able to be classified on the basis of plant status alone (e.g., fuel handling accident in spent fuel building).

The actual or projected dose of 100 mR TEDE is set at 10% of the EPA Protective Action Guide (PAG) values given in EPA-400-R-92-001, while the 500 mR child thyroid CDE was established in consideration of the 1:5 ratio of the EPA PAG for TEDE and thyroid CDE. The TEDE integrated dose value also provides a desirable gradient (one order of magnitude) between the Alert, Site Area Emergency and General Emergency Classes.

The SM/TSC Director/EOF Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 15 minutes.

The monitor list inmEAL #1 includes monitors on all potential release pathways (plant stack, primary-secondary leak, fuel handling accident). The EPA PAGs are expressed in terms of the sum of the "effective dose equivalent (EDE)" and the "committed effective dose equivalent (CEDE)", or as the child thyroid "committed dose equivalent (CDE)". For the purpose of these ICs, the dose quantity "total effective dose equivalent (TEDE)", as defined in 10 CFR 20, is used in lieu of "...sum of EDE and CEDE...." The EPA PAG guidance in EPA-40OR-92-001 provides for the use adult thyroid dose conversion factors.

The monitor readings in EAL #1 were determined using a dose assessment method that back-calculates from the dose values specified. The meteorology and source term (noble gases, particulates, and halogens) used are the same as those used for determining the monitor reading EALs in AU1 and AA1. This protocol maintains intervals between the ICs for the four classifications.

Since doses are generally not monitored in real-time, a release duration of one hour was assumed, and the EALs are based on a site boundary (or beyond) dose of 100 mR/hour whole body or 500 mR/hour child thyroid, whichever is more limiting (as was done for EAL #3).

Monitor indications in EAL #1 are calculated using SAR source terms applicable to each monitored pathway in conjunction with annual average meteorology, one hour release duration and Dose Conversion Factors (DCFS) from EPA-40OR-92-001, Tables 5-1 and 5-2.

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/WORK PLAN TITLE: PAGE: 51 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT (AS1)

Since dose assessment in EAL #2 is based on actual meteorology, whereas the monitor readings in EAL #1 are not, the results from these assessments may indicate that the classification is not warranted, or may indicate that a higher classification is warranted. For this reason, emergency implementing procedures should call for performance of dose assessments within 15 minutes using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g., initiated at a lower classification level), the dose assessment results override the monitor reading EALs. However, classification should not be delayed pending the results of these dose assessments. If dose assessment team calculations cannot be completed in 15 minutes, then valid monitor readings should be used for emergency classification.

Field team surveys in EAL #3 should be performed at or beyond the SITE BOUNDARY and at the most accurate indicator of the condition. Field data are independent of release elevation and meteorology. The assumed release duration is one hour. Expected post accident source terms would be. dominated by noble gases providing the dose rate value. Sampling of radioiodine by adsorption on a charcoal cartridge should determine the iodine value.

Escalation is via AG1.

PROCJWORK PLAN NO. PROCEDUREJWORK PLAN TITLE: PAGE: 52 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-X(X-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AGI GENERAL EMERGENCY Initiating Condition:

Offsite dose resulting from an actual or imminent release of gaseous radioactivity exceeds 1000 mR TEDE or 5000 mR child thyroid CDE for the actual or projected duration of the release using actual meteorology Operating Mode Applicability:

All Emergency Action Level: 1 OR 2 OR 3 Note: If dose assessment results are available at the time of declaration, the classification should be based on EAL #2 instead of EAL #1. While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated/completed in order to determine if the classification should be subsequently escalated.

1. VALID reading on Channel 9 of one or more of the following radiation monitors that exceeds or is expected to exceed the reading shown for 15 minutes or longer:

MONITORS - UNIT 1 LIMIT RX-9820 Containment Purge 5.90E+2 (pCi/cc)

RX-9830 Fuel Handling Area 4.54E+2 (pCi/cc)

RX-9825 Radwaste Area 5.36E+2 (pCi/cc)

RX-9835. Emergency Penetration Room 9.56E+3 (pCi/cc)

LIMIT MONITORS - UNIT 2 2RX-9820 Containment Purge 4.46E+2 (pCi/cc) 2RX-9825 Radwaste Area 3.32E+2 (pCi/cc) 2RX-9830 Fuel Handling Area. 4.46E+2 (pCi/cc) 2RX-9835 Emergency Penetration Room 8.84E+3 (pCi/cc) 2RX-9840 Post Accident Sampling Building 4.42E+3 (pCi/cc) 2RX-9845 Aux. Building Extension 1.26E+3 (pCi/cc) 2RX-9850 Low Level Radwaste Storage Building. 1.77E+3 (pCi/cc)

2. Dose assessment using actual meteorology indicates doses greater than 1000 mR TEDE or 5000 mR child thyroid CDE at or beyond the site boundary.

3. Field survey results indicate closed window dose rates exceeding 1000 mR/hr expected to continue for more than one hour; or analyses of field survey samples indicate child thyroid CDE of 5000 mR for 60 minutes of inhalation, at or beyond site boundary.

PROC.MWORK PLAN NO. PROCEDUREINORK PLAN TITLE: PAGE: 53 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT (AGI)

Basis:

This IC and associated EALs address radioactivity releases that result in doses at or beyond the site boundary that exceed the EPA Protective Action Guides (PAGs). Public protective actions will be necessary. Releases of this magnitude are associated with the failure of plant systems needed for the protection of the public and likely involve fuel damage. While these failures are addressed by other EALs, this EAL provides appropriate diversity and addresses events which may not be able to be classified on the basis of plant status alone. It is important to note that, for the more severe accidents, the release may be unmonitored-or there may be large uncertainties associated with the source term and/or meteorology.

The actual or projected dose of 1000 mR TEDE and 5000 mR child thyroid CDE integrated doses are based on the EPA Protective Action Guide (PAG) values given in EPA-400-R-92-001, which indicates that public protective actions are indicated if doses exceed these values. This is consistent with the emergency class description of a General Emergency.

The SM/TSC Director/EOF Director should not wait until 15 minutes has elapsed, but should declare the event as soon as it is determined that the release duration has or will likely exceed 15 minutes.

The monitor list in EAL #l includes monitors on all potential release pathways (Plant stack, Primary/Secondary Leak, Fuel Handling Accident). The EPA PAGs are expressed in-terms of the sum of the "effective dose equivalent (EDE)" and the "committed effective dose equivalent (CEDE)", or as the child thyroid "committed dose equivalent (CDE)". For the purpose of these ICs, the dose quantity "total effective dose equivalent (TEDE)", as defined in 10 CFR 20, is used in lieu of "...sum of EDE and CEDE...." The EPA PAG guidance EPA-400R-92-001 provides for the use of adult thyroid dose conversion factors.

The monitor readings in EAL #1 were determined using a dose assessment method that back-calculates from the dose values. The meteorology and source term (noble gases, particulates, and halogens) used are the same as those used for determining the monitor reading EALs in AU1 and AA1. This protocol maintains intervals between the EALs for the four classifications. Since doses are generally not monitored in real-time, a release duration of one hour was assumed, and the ICs are based on a site boundary (or beyond) dose of 1000 mR/hour whole body or 5000 mR/hour child thyroid, whichever is more limiting (as was done for EAL #3). If the site analyses indicate a longer or shorter duration for the period in which the substantial portion of the activity is.

released, the longer duration should be used.

Monitor indications in EAL #1 are calculated using SAR source terms applicable to each monitored pathway in conjunction with annual average meteorology, one hour release duration and dose conversion factors (DCFs) from EPA-400R-92-001, Tables 5-1 and 5-2.

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 54 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT (AGI)

Since dose assessment in EAL #2 is based on actual meteorology, whereas the monitor reading in EAL #1 are not, the results from these assessments may indicate that the classification is not warranted. For this reason, emergency implementing procedures should call for performance of dose assessments within 15 minutes using actual meteorology and release information. If the results of these dose assessments are available when the classification is made (e.g.,

initiated at a lower classification level), the dose assessment results override the monitor reading EALs. However, classification should not be delayed pending the results of these dose assessments. If dose assessment team calculations cannot be completed in 15 minutes, then valid monitor readings should be used for emergency classification.

Field team surveys in EAL #3 should be performed at or beyond the SITE BOUNDARY and at the most accurate indicator of the condition. Field data are independent of release elevation and meteorology. The assumed release duration is one hour. Expected post accident source terms would be dominated by noble gases providing the dose rate value. Sampling of radioiodine by adsorption on charcoal cartridge should determine the iodine value.

PROC.JWORK PLAN NO. PROCEDUREJWORK PLAN TITLE: PAGE: 55 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction cul NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

RCS leakage Operating Mode Applicability:

Cold Shutdown (Mode 5)

Emergency Action Level: 1 OR 2

1. Unidentified or pressure boundary leakage greater than 10 gpm.

2. Identified leakage greater than 25 gpm.

Basis:

This IC is included as an NUE because it is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified and pressure boundary leakage was selected as it is sufficiently large to be observable via normally installed instrumentation (e'g.,

pressurizer level, RCS loop level instrumentation, etc.) or reduced inventory instrumentation such as level hose indication. Lesser values must generally be determined through time consuming surveillance tests (e.g., mass balances).

The EAL for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage. Prolonged loss of RCS inventory may result in escalation to the ALERT level via either IC CA1 or CA4.

The difference-between CUl and CU2 deals with the RCS conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown the RCS will normally be intact and RCS inventory and level monitoring-means such as pressurizer level indication and makeup volume control tank levels are normally available.

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 56 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction CU2 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

UNPLANNED loss of RCS inventory with irradiated fuel in the reactor vessel Operating Mode Applicability:

Refueling (Mode 6)

Emergency Action Level: 1 OR 2

1. UNPLANNED RCS level drop below the reactor vessel flange for greater than 15 minutes.

2. a. Loss of reactor vessel inventory as indicated by unexplained Reactor Building Sump, Reactor Drain Tank, Aux. Building Equipment Drain Tank, Aux. Building Sump, or Quench Tank level rise.

AND

b. Reactor vessel level cannot be monitored.

Basis:

This IC is included as an NUE because it may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. Refueling evolutions that lower RCS water level below the reactor vessel flange are carefully planned and procedurally controlled. An UNPLANNED event that results in water level dropping below the reactor vessel flange warrants declaration of an NUE due to the reduced RCS inventory that is available to keep the core covered. The allowance of 15 minutes was chosen because it is reasonable to assume that level can be restored within this time frame using one or more of the redundant means of refill that should be available. If level cannot be restored in this time frame, then it may indicate that a more serious condition exists. Continued loss of RCS inventory will result in escalation to the ALERT level via either IC CA2 or CA4.

The difference between CUI and CU2 deals with the RCS conditions that exist between cold shutdown and refueling modes. In cold shutdown the RCS will normally be intact and standard indications of RCS inventory are available.

In the refueling mode, normal means of core temperature indication and RCS level indication may not be available. Redundant means of reactor vessel level indication will normally be installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted.

However, if all level indication were to be lost during a loss of RCS inventory

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/WORK PLAN TITLE: PAGE: 57 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction (CU2) event, the operators would need to determine that reactor vessel inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. Escalation to ALERT would be via either CA2 or CA4.

EAL #1 involves a drop in RCS level below the top of the reactor vessel flange that continues for 15 minutes due to an UNPLANNED event. This EAL is not applicable to lowering levels in flooded refueling canal level (covered by AU2, EAL iU) until such time as the level lowering to the level of the vessel flange.

If the reactor vessel level continues to lower and reaches the bottom of the reactor coolant system hot leg penetration into the vessel, then escalation to CA2 would be appropriate.

PROC.JORK PLAN NO. PROCEDURENWORK PLAN TITLE: PAGE: 58 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 .

Attachment 3 Cold Shutdown/Refueling System Malfunction CU3 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Loss of all offsite power to vital busses for greater than 15 minutes Operating Mode Applicability:

Cold Shutdown (Mode 5)

Refueling (Mode 6)

Defueled Emergency Action Level:

1. a. Loss of power to all unit auxiliary and startup transformers supplying a unit for greater than 15 minutes.

AND

b. At least one vital 4.16 KV bus being powered from ANY diesel generator.

Basis:

Prolonged loss of AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete loss of AC Power (e.g., station blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalation is via CA3.

PROC.IWORK PLAN NO. PROCEDURENVORK PLAN TITLE: PAGE: 59 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction CU4 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

UNPLANNED loss of decay heat removal capability with irradiated fuel in the reactor vessel Operating Mode Applicability:

Cold Shutdown (Mode 5)

Refueling (Mode 6)

Emergency Action Level: 1 OR 2

1. An UNPLANNED event results in RCS temperature exceeding 2000F.

2. Loss of all RCS temperature and reactor vessel level indication for greater than 15 minutes.

Basis:

This IC is included as an NUE because it may be a precursor of more serious conditions and, as a result, is considered to be a potential degradation of the level of safety of the plant. In cold shutdown the ability to remove decay heat relies primarily on forced cooling flow. Operation of the systems that provide this forced cooling may be jeopardized due to the unlikely loss of electrical power or RCS inventory. Since the RCS usually remains intact in the cold shutdown mode, a large inventory of water is available to keep the core covered. In cold shutdown, the decay heat available to raise RCS temperature during a loss of inventory or loss of heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown conditions may be attained within hours of operating at power. Entry into the refueling mode procedurally may' not occur for many hours after the reactor has been shut down.

Thus, the heatup threat (and, therefore, the threat to damaging the fuel clad) may be lower for events that occur in the refueling mode with irradiated fuel in the reactor vessel. In addition, the operators should be able to monitor RCS temperature and reactor vessel level so that escalation to the ALERT level via CA2 or CA4 will occur if required.

Loss of forced decay heat removal at reduced inventory may result in more rapid rises in reactor coolant temperatures depending on the time since-shutdown.

Escalation to the Alert level via CA4 is provided should an UNPLANNED event result in RCS temperature exceeding the Technical Specification cold shutdown temperature limit for greater than 30 minutes with CONTAINMENT CLOSURE not established.

Unlike the cold shutdown mode, normal means of core temperature indication and RCS levelindication may not be available in the refueling mode. Redundant means of reactor vessel level indication are procedurally installed to assure that the ability to monitor level will not be interrupted. However, if all

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/WORK PLAN TITLE: PAGE: 60 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction (CU4) level and temperature indication were to be lost in.either the cold shutdown or refueling modes, EAL #2 would result in declaration of an NUE if either temperature or level indication cannot be restored within 15 minutes from the loss of both means of indication. Escalation to ALERT would be via CA2 based on an inventory loss or CA4 based on exceeding its temperature criterion.

The SM/TSC Director/EOF Director must remain attentive to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment of the SM/TSC Director/EOF Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded.

PROCJWORK PLAN NO. PROCEDUREMORK PLAN TITLE: PAGE: 61 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction CU5 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Fuel clad degradation Operating Mode Applicability:

Cold-Shutdown (Mode 5)

Refueling (Mode 6)

Emergency Action Level:

1. RCS sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits.

Unit 1:

Greater than 3.50 pCi/gm IDE Greater than 72/E pCi/gm Gross Activity Unit 2:

Greater than 1.0 pCi/gm IDE Greater than 100/E pCi/gm Gross Activity Basis:

The condition noted in this EAL is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. This EAL addresses reactor coolant samples exceeding Technical Specifications for iodine spikes that are indicative of a loss of fuel clad integrity.

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 62 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction CU6 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

UNPLANNED loss of all onsite or offsite communications capabilities Operating Mode Applicability:

Cold Shutdown (Mode 5)

Refueling (Mode 6)

Emergency Action Level: 1 OR 2

1. Loss of all onsite communications capability (Table Cl) affecting the ability to perform routine operations.

Table Cl Onsite Communications Equipment Station radio system Plant paging system In-plant telephones Plant cell phones Gaitronics

2. Loss of all offsite communications capability (Table C2).

Table C2 Offsite Communications Equipment All telephone lines (commercial and microwave)

Station radio system ENS Cellular phones Basis:

The purpose of this IC and its associated EALs is to recognize a loss of communications capability that either defeats the plant operations staff's ability to perform routine tasks necessary for plant operations or the ability to communicate problems to offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72. The availability of one method of ordinary offsite communications is sufficient to inform state and local authorities of plant problems. This EAL is intended to be used only when extraordinary means (e.g., relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being utilized to make communications possible.

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 63 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction CU7 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

UNPLANNED loss of required DC power for greater than 15 minutes Operating Mode Applicability:

Cold Shutdown (Mode 5)

Refueling (Mode 6)

Emergency Action Level:

1. a. UNPLANNED Loss of Vital DC power to required DC busses based on bus voltage indicating 105 volts or less.

AND

b. Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.

Basis:

The purpose of this IC and its associated EALs is to recognize a loss of DC power compromising the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations. This EAL is intended to be anticipatory since the operating crew may not have necessary indication and control of equipment needed to respond to the loss.

UNPLANNED is included in this IC and EAL to preclude the declaration of an emergency as a result of planned maintenance activities. Routinely, plants perform maintenance on a train related basis during shutdown periods. It is intended that the loss of the operating (operable) train is to be considered.

The specified bus voltage indication, 105 volts, is based on the minimum bus voltage necessary for the operation of safety related equipment.

If the loss of DC power results in the inability to maintain cold shutdown, the escalation to an ALERT will be per CA4.

PROC.MWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 64 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0.

Attachment 3 Cold Shutdown/Refueling System Malfunction CU8 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Inadvertent criticality Operating Mode Applicability:

Cold Shutdown (Mode 5)

Refueling (Mode 6)

Emergency Action Level:

1. An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.

Basis:

This IC addresses criticality events that occur in cold shutdown or refueling modes (NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States) such as fuel misloading events and inadvertent dilution events. This condition indicates a potential degradation of the level of safety of the plant warranting an NUE classification. The IC excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated) which are addressed in the companion IC SU8.

This condition can be identified using the startup rate monitor. The term "sustained" is used in order to allow exclusion of expected short term positive startup rates from planned fuel bundle or control rod movements during core alteration. These short term positive startup rates are the result of therise in neutron population due to subcritical multiplication.

Escalation would be by SM/TSC Director/EOF Director judgment.

PROCJWORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 65 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction CA1 ALERT Initiating Condition:

Loss of RCS inventory Operating Mode Applicability:

Cold.Shutdown (Mode 5)

Emergency Action Level:

1. a. Loss of reactor vessel inventory as indicated by unexplained Reactor Building Sump, Reactor Drain Tank, Aux. Building Equipment Drain Tank, Aux. Building Sump, or Quench Tank level rise.

AND

b. RCS level cannot be monitored for greater than 15 minutes.

Basis:

This EAL serves as a precursor to a loss of heat removal. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further reactor vessel level drop and potential core uncovery. This condition will result in a minimum classification of ALERT. The bottom of the RCS hot leg penetration into the reactor vessel is 368 ft., 0 in. (Unit 1) or 369 ft., 1.5 in. (Unit 2). Below this level, remote RCS level indication may be lost and loss of suction to decay heat removal systems may occur. The inability to restore and maintain level after reaching this setpoint would, therefore, be indicative of a failure of the RCS barrier.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for several hours after the reactor has been shutdown. Thus the heatup threat and therefore the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the reactor vessel. The above forms the basis for needing both a cold shutdown specific IC (CA1) and a refueling specific IC (CA2).

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INORK PLAN TITLE: PAGE: 66 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-O Attachment 3 Cold Shutdown/Refueling System Malfunction (CAI)

In the refueling mode, normal means of reactor vessel level indication may not be available. Redundant means of reactor vessel level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indications were to be lost during a loss of RCS inventory event, the operators would need to determine that reactor vessel inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of reactor vessel leakage. The 15-minute duration for the loss of level indication was chosen because it is half of the CS2 Site Area Emergency EAL duration. The 15-minute duration allows CA2 to be an effective precursor to CS2. Significant fuel damage is not expected to occur until the core has been uncovered for greater than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per the analysis referenced in the CS2 basis. Therefore, this EAL meets the definition for an ALERT.

The difference between CAl and CA2 deals with the reactor conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown the reactor vessel will normally be intact and standard reactor vessel level monitoring means are available.

If reactor vessel level continues to drop, then escalation to Site Area Emergency will be via CS1.

PROC.INORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 67 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXXXX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction CA2 ALERT Initiating Condition:

Loss of reactor vessel inventory with irradiated fuel in the reactor vessel Operating Mode Applicability:

Refueling (Mode 6)

Emergency Action Level:

1. a. Loss of reactor vessel inventory as indicated by unexplained Reactor Building Sump, Reactor Drain Tank, Aux. Building Equipment Drain Tank, Aux. Building Sump, or Quench Tank level rise.

AND

b. Reactor vessel level cannot be monitored for greater than 15 minutes.

Basis:

This EAL serves as a precursor to a loss of heat removal. The magnitude of this loss of water indicates that makeup systems have not been effective and may not be capable of preventing further reactor vessel level drop and potential core uncovery. This condition will result in a minimum classification of ALERT. The bottom of the RCS hot leg penetration into the reactor vessel is 368 ft., 0 in. (Unit 1) or 369 ft., 1.5 in. (Unit 2). Below this level, remote RCS level indication may be lost and loss of suction to decay heat removal systems may occur. The inability to restore and maintain level after reaching this setpoint would, therefore, be indicative of a failure of the RCS barrier.

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for several hours after the reactor has been shutdown. Thus, the heatup threat and, therefore, .the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the reactor vessel. The above forms the basis for needing both a cold shutdown specific IC (CA1) and a refueling specific IC (CA2).

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/WORK PLAN TITLE: PAGE: 68 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction (CA2)

In the refueling mode, normal means of reactor vessel level indication may not be available. Redundant means of reactor vessel level indication will be normally installed (including the ability to monitor level visually) to assure that the ability to monitor level will not be interrupted. However, if all level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that reactor vessel inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of reactor vessel leakage. The 15-minute duration for the loss-of level indication was chosen because it is half of the CS2 Site Area Emergency EAL duration. The 15-minute duration allows CA2 to be an effective precursor to CS2. Significant fuel damage is not expected to occur until the core has been uncovered for greater than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per the analysis referenced in the CS2 basis. Therefore, this EAL meets the definition for an ALERT.

The difference between CA1 and CA2 deals with the reactor conditions that exist between cold shutdown and refueling mode applicability. In cold shutdown the reactor vessel will normally be intact and standard reactor vessel level monitoring means are available. In the refueling mode the reactor vessel is not intact and reactor vessel inventory is monitored by different means.

If reactor vessel level continues to drop, then escalation to Site Area Emergency will be via CS2.

PROC.MORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 69 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction CA3 ALERT Initiating Condition:

Loss of all offsite power and loss of all onsite AC power to required 4.16 KV busses Operating Mode Applicability:

Cold Shutdown (Mode 5)

Refueling (Mode 6)

Defueled Emergency Action Level:

1. a. Loss of power to all unit auxiliary and startup transformers supplying a unit.

AND

b. No diesel generator is supplying power to emergency busses on the affected unit.

AND

c. Failure to restore power to at least one emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power.

Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including DHR/shutdown cooling, emergency core cooling, containment cooling, spent fuel pool cooling, and the ultimate heat sink. When in the cold shutdown, refueling, or defueled mode the event can be classified as an Alert because of the significantly reduced decay heat and lower temperature and pressure which allow raising the time to restore one of the emergency busses, relative to that specified for the Site Area Emergency EAL. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

Escalating to Site Area Emergency, if appropriate, is by Abnormal Rad Levels/Radiological Effluent, or SM/TSC Director/EOF Director judgment ICs.

Consideration should be given to available loads necessary to remove decay heat or provide reactor vessel makeup capability when evaluating loss of AC power to vital busses. Even though a vital bus may be energized, if necessary loads (i.e., loads that, if lost, would inhibit decay heat removal capability or reactor vessel makeup capability) are not available on the energized bus, then the bus should not be considered available.

PROCJWORK PLAN NO. PROCEDUREINORK PLAN TITLE: PAGE: 70 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction CA4 ALERT Initiating Condition:

Inability to maintain plant in cold shutdown with irradiated fuel in the reactor vessel Operating Mode Applicability:

Cold Shutdown (Mode 5)

Refueling (Mode 6)

Emergency Action Level: 1 OR 2 OR 3

1. With CONTAINMENT CLOSURE and RCS integrity not established, an UNPLANNED event results in RCS temperature exceeding 2000 F

2. With CONTAINMENT CLOSURE established AND RCS integrity not established OR RCS inventory reduced, an UNPLANNED event results in RCS temperature exceeding 200'F for greater than 20 minutes'.

3. An UNPLANNED event results in RCS temperature exceeding 200'F for greater than 60 minutes' or results in an RCS pressure rise of greater than 10 psi.

'Note: IF decay heat removal system (Decay Heat or Shutdown Cooling) is in operation within this time frame AND RCS temperature is being reduced, THEN this EAL is not applicable.

Basis:

This IC and its associated EALs are based on concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal. A number of phenomena such as pressurization, vortexing, steam generator U-tube draining, RCS level differences when operating at a mid-loop condition, decay heat removal system design, and level instrumentation problems can lead to conditions where decay heat removal is lost and core uncovery can occur. NRC analyses show that sequences of events can cause core uncovery in 15 to 20 minutes and severe core damage within an hour after decay heat removal is lost.

A loss of Technical Specification components alone is not intended to constitute an Alert. The same is true of a momentary UNPLANNED excursion above 200 0 F when the heat removal function is available.

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MWORK PLAN TITLE: PAGE: 71 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-O Attachment 3 Cold Shutdown/Refueling System Malfunction (CA4)

The SM/TSC Director/EOF Director must remain alert to events or conditions that lead to the conclusion that exceeding the EAL threshold is imminent. If, in the judgment .of the SM/TSC Director/EOF Director, an imminent situation is at hand, the classification should be made as if the threshold has been exceeded.

EAL #1 addresses complete loss of functions required for core cooling during refueling and cold shutdown modes when neither CONTAINMENT CLOSURE nor RCS integrity are established. RCS integrity is in place when the RCS pressure boundary is in its normal condition to be pressurized (e.g., no freeze seals or nozzle dams). No delay time is allowed for EAL #1 because the evaporated reactor coolant that may be released into the containment during this heatup condition could also be directly released to the environment.

EAL #2 addresses the complete loss of functions required for core cooling for greater than 20 minutes during refueling and cold shutdown modes when CONTAINMENT CLOSURE is established but RCS integrity is not established or RCS inventory is reduced (e.g., mid-loop operation). As in EAL #1, RCS integrity should be assumed to be in place when the RCS pressure boundary is in its normal condition to be pressurized (e.g., no freeze seals or nozzle dams). The allowed 20-minute time frame was included to allow operator action to restore the heat removal function, if possible. The allowed time frame is consistent with the guidance provided by Generic Letter 88-17, Loss of Decay Heat Removal, and is believed to be conservative given that a low pressure containment barrier to fission product release is established. Note 1 indicates that EAL #2 is not applicable if. actions are successful in restoring an RCS heat removal system to operation and RCS temperature is being reduced within the 20-minute time frame.

EAL #3 addresses complete loss of functions required for core cooling for greater than 60 minutes during refueling and cold shutdown modes when RCS integrity is established. As in EAL #1 and #2, RCS integrity should be considered to be in place when the RCS pressure boundary is in its normal condition to be pressurized. (e.g., no freeze seals or nozzle dams). The status of CONTAINMENT CLOSURE in this EAL is immaterial given that the RCS is providing a high pressure barrier to fission product release to the environment. The 60-minute time frame should allow sufficient time to restore cooling without a substantial degradation in plant safety. The 10 psi pressure rise covers situations where, due to high decay heat loads, the time provided to restore temperature control should be less than 60 minutes. The RCS pressure setpoint chosen is 10 psi, which-can be read on installed control board instrumentation. Note 1 indicates that EAL 3 is not applicable if actions are successful in restoring a shutdown cooling system to operation and RCS temperature is being reduced within the 60-minute time frame assuming that the RCS pressure rise has remained less than 10 psi.

Escalation to Site Area Emergency would be via CS1 or CS2 should boiling result in significant reactor vessel level loss leading to core uncovery.

PROCJWORK PLAN NO. PROCEDURENVORK PLAN TITLE: PAGE: 72 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction CS1 SITE AREA EMERGENCY Initiating Condition:

Loss of reactor vessel inventory affecting core decay heat removal capability Operating Mode Applicability:

Cold Shutdown (Mode 5)

Emergency Action Level: 1 OR 2

1. With CONTAINMENT CLOSURE not established:

a. Loss of reactor vessel inventory is indicated by unexplained Reactor Building Sump, Reactor Drain Tank, Quench Tank, Aux. Building Equipment Drain Tank, or Aux. Building Sump level rise.

AND

b. Reactor vessel level cannot be monitored for greater than 30 minutes.

2. With CONTAINMENT CLOSURE established:

a. Loss of reactor vessel inventory is indicated by either:

Unexplained Reactor Building Sump, Reactor Drain Tank, Quench Tank, Aux. Building Equipment Drain Tank, or Aux. Building Sump level rise

Erratic source range monitor indication

Core exit thermocouples indicating superheat AND

b. Reactor vessel level cannot be monitored for greater than 30 minutes.

Basis:

Under the conditions specified by these EALs, continued lowering in reactor vessel level is indicative of a loss of inventory control. Inventory loss may be due to a reactor vessel breach, pressure boundary leakage, or continued boiling in the reactor vessel.

PROCJWORK PLAN NO.' PROCEDURE/WORK PLAN TITLE: PAGE: 73 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-O Attachment 3 Cold Shutdown/Refueling System Malfunction (CS1)

If all reactor vessel level indications were to be lost during a loss of RCS inventory event, the operators would need to determine that reactor vessel inventory loss was occurring by observing containment sump level, reactor drain tank level, or quench tank level change. Containment sump level, reactor drain tank level, or quench tank level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. This EAL is based on concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal, SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues, NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States, and, NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. A number of variables, (mid-loop, reduced level/flange level, head in place, cavity flooded, RCS venting strategy, decay heat removal system design, vortexing pre-disposition, or steam generator U-tube draining) can have a significant impact on heat removal capability challenging the fuel clad barrier. Analysis in the above references indicates that core damage may occur within an hour following continued core uncovery; therefore, 30 minutes was chosen to be conservative.

The 30-minute duration allowed when CONTAINMENT CLOSURE is established allows sufficient time for actions to be performed to recover needed cooling equipment and is considered to be conservative. As water level in the reactor vessel lowers, the dose rate above the core will rise. Additionally, studies indicate that the installed nuclear instrumentation will operate erratically when the core is uncovered and can be used as a tool for making such determinations. In the refueling mode, normal means of reactor vessel level indication may not be available; however, redundant means of reactor vessel level indication is normally installed to assure that the ability to monitor level will not be interrupted. Since effluent release is not expected with closure established, declaration of a Site Area Emergency is warranted under the conditions specified.

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 74 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction CS2 SITE AREA EMERGENCY Initiating Condition:

Loss of reactor vessel inventory affecting core decay heat removal capability with irradiated fuel in the reactor vessel Operating Mode Applicability:

Refueling (Mode 6)

Emergency Action Level: 1 OR 2

1. With CONTAINMENT CLOSURE not established:

a. Loss of reactor vessel inventory is indicated by unexplained Reactor Building Sump, Reactor Drain Tank, Quench Tank, Aux. Building Equipment Drain Tank, or Aux. Building Sump level rise.

b. Reactor vessel level cannot be monitored for greater than 30 minutes.

2. With CONTAINMENT CLOSURE established:

a. Loss of reactor vessel inventory is indicated by either:

Unexplained Reactor Building Sump, Reactor Drain Tank, Quench Tank, Aux Building Equipment Drain Tank, or Aux Building Sump level rise

Erratic source range monitor indication

Core exit thermocouples indicating superheat AND

b. Reactor vessel level cannot be monitored for greater than 30 minutes.

Basis:

Under the conditions specified by these EALs, continued drop in reactor vessel level is indicative of a loss of inventory control. Inventory loss may be due to a reactor vessel breach, pressure boundary leakage, or continued boiling in the reactor vessel.

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MWORK PLAN TITLE: PAGE: 75 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction (CS2)

In cold shutdown the decay heat available to raise RCS temperature during a loss of inventory or heat removal event may be significantly greater than in the refueling mode. Entry into cold shutdown conditions may be attained within hours of operating at power or hours after refueling is completed. Entry into the refueling mode procedurally may not occur for several hours after the reactor has been shutdown. Thus the heatup threat and, therefore, the threat to damaging the fuel clad may be lower for events that occur in the refueling mode with irradiated fuel in the reactor vessel (note that the heatup threat could be lower for cold shutdown conditions if the entry into cold shutdown was following a refueling). The above forms the basis for needing both a cold shutdown specific IC (CS1) and a refueling specific IC (CS2)

If all reactor vessel level indication were to be lost during a loss of RCS inventory event, the operators would need to determine that reactor vessel inventory loss was occurring by observing containment sump level, reactor drain tank level, or quench tank level change. Containment sump level, reactor drain tank level, or quench tank level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage. This EAL is based on concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal, SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues, NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States, and, NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. A number of variables, (mid-loop, reduced level/flange level, head in place, cavity flooded, RCS venting strategy, decay heat removal system design, vortexing pre-disposition, or steam generator U-tube draining) can have a significant impact on heat removal capability challenging the fuel clad barrier. Analysis in the above references indicates that core damage may occur within an hour following continued core uncovery; therefore, 30 minutes was chosen to be conservative.

The 30-minute duration allowed when CONTAINMENT CLOSURE is established allows sufficient time for actions to be performed to recover needed cooling equipment and is considered to be conservative. As water level in the reactor vessel lowers, the dose rate above the core will rise. Additionally, studies indicate that the installed nuclear instrumentation will operate erratically when the core is uncovered and can be used as a tool for making such determinations. In the refueling mode, normal means of reactor vessel level indication may not be available; however, redundant means of reactor vessel level indication is normally installed to assure that the ability to monitor level will not be interrupted. Since effluent release is not expected with closure established, declaration of a Site Area Emergency is warranted under the conditions specified.

Declaration of an Site Area Emergency is warranted under the conditions specified by the IC. Escalation to a General Emergency is via CG1 or AG1.

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 76 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction CG1 GENERAL EMERGENCY Initiating Condition:

Loss of reactor vessel inventory affecting fuel clad integrity with containment challenged with irradiated fuel in the reactor vessel Operating Mode Applicability:

Cold Shutdown (Mode 5)

Refueling (Mode 6)

Emergency Action Level: 1 AND 2 AND 3

1. Loss of reactor vessel inventory as indicated by unexplained Reactor Building Sump, Reactor Drain Tank, Quench Tank, Aux. Building Equipment Drain Tank, or Aux. Building Sump level rise.

2. Reactor vessel level cannot be monitored for greater than 30 minutes with indication of core uncovery, as evidenced by one or more of the following:

Erratic source range monitor indication.

Core exit thermocouples indicating superheat.

3. Indication of CONTAINMENT challenged as indicated by one or more of the following:

Containment hydrogen greater than or equal to 4%.

Pressure above 59 psig (Unit 1) or 73.7 psia (Unit 2) with CONTAINMENT INTEGRITY.

CONTAINMENT CLOSURE not established.

Basis:

For EAL #1 the operators would need to determine that reactor vessel inventory loss was occurring by observing sump and tank level changes. Sump and tank level rises must be evaluated against other potential sources of leakage such as cooling water sources inside the containment to ensure they are indicative of RCS leakage.

PROC.MVORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 77 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction (CGI)

EAL #2 represents the inability to restore and maintain reactor vessel level above the top of active fuel. Fuel damage is probable if reactor vessel level cannot be restored, as available decay heat will cause boiling further reducing the reactor vessel level. These EALs are based on concerns raised by Generic Letter 88-17, Loss of Decay Heat Removal, SECY 91-283, Evaluation of Shutdown and Low Power Risk Issues, NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States, and NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management. A number of variables (e.g., mid-loop, reduced level/flange level, head in place, cavity flooded, RCS venting strategy, decay heat removal system design, vortexing pre-disposition, or steam generator U-tube draining) can have a significant impact on heat removal capability challenging the fuel clad barrier. Analysis in the above references indicates that core damage may occur within an hour following continued core uncovery; therefore, 30 minutes was chosen to be conservative.

As water level in the reactor vessel lowers, the dose rate above the core will rise. Additionally, post-TMI studies indicated that the installed nuclear instrumentation will operate erratically when the core is uncovered and that this should be used as a tool for making such determinations. The GE is declared on the occurrence of the loss or imminent loss of function of all three barriers. Based on the above discussion, RCS barrier failure resulting in core uncovery for 30 minutes or more may cause fuel clad failure. With the CONTAINMENT breached or challenged, the potential for unmonitored fission product release to the environment is high. This represents a direct path for radioactive inventory to be released to the environment. This is consistent with the definition of a GE.

In the context of EAL #3, containment closure is the action taken to secure containment and its associated structures, systems, and components as a functional barrier to fission product release under existing plant conditions.

Containment closure should not be confused with refueling containment integrity as defined in technical specifications. Site shutdown contingency plans provide for re-establishing containment closure following a loss of heat removal or RCS inventory functions. If the closure is re-established prior to exceeding the temperature or level thresholds of the RCS barrier and fuel clad barrier EALs, escalation to GE would not occur.

The pressure at which containment is considered challenged is based on the condition of the containment. If containment integrity is established, then the containment will be challenged at the design pressure. This is consistent with the owners groups' Emergency Response Procedures. Since no significant pressurization is expected during cold shutdown/refueling operations, there is no specific pressure setpoint at which the containment is considered to be challenged. Plant procedures provide for the establishment of containment closure when required and for the monitoring of the status of containment closure.

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/WORK PLAN TITLE: PAGE: 78 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Cold Shutdown/Refueling System Malfunction (CG1)

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 mixture of dissolved gasses in containment. However, containment monitoring and/or sampling should be performed to verify that hydrogen concentrations greater than 4.0% exist. The 4.0% value, though not representative of an explosive mixture of hydrogen, is consistent with the concentration that can be maintained with at least on hydrogen recombiner in service.

PROC.JWORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 79 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 EVENTS RELATED TO ISFSI E-HU1 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Damage to a loaded cask CONFINEMENT BOUNDARY Operating Mode Applicability:

Not Applicable Emergency Action Level: 1 OR 2 OR 3

1. Natural phenomena events affecting.a loaded cask CONFINEMENT BOUNDARY.

Tornado/High winds

Flood

2. Accident conditions affecting a loaded cask CONFINEMENT BOUNDARY.

Cask drop accident

Blockage of air inlets

Fire or explosion

3. Any condition in the opinion of the SM/TSC Director/EOF Director that indicates loss of loaded fuel storage cask CONFINEMENT BOUNDARY.

Basis:

A NUE would be declared on the basis of the occurrence of an event of sufficient magnitude that a loaded cask CONFINEMENT BOUNDARY is damaged or violated. This includes classification based on a loaded fuel storage cask CONFINEMENT BOUNDARY loss leading to the degradation of the fuel during storage or posing an operational safety problem with respect to its removal from storage.

For EAL #1 and EAL #2, the results of the ISFSI Safety Analysis Report (SAR) referenced in the cask('s) Certificate of Compliance and the related NRC Safety Evaluation Report are used to develop a list of natural phenomena events and accident conditions. These EALs address responses to a dropped cask, a tipped-over cask, explosion, missile damage, fire damage or natural phenomena affecting a cask (e.g., seismic event, tornado, etc.).

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 80 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 EVENTS RELATED TO ISFSI (E-HU1)

For EAL #3, any condition not explicitly detailed as an EAL threshold value, which, in the .judgment of the SM/TSC Director/EOF Director, is a potential degradation in the level of safety of the ISFSI. SM/TSC Director/EOF Director judgment is to be based on known conditions and the expected response to mitigating activities within a short time period.

Possible damage mode to the storage.cask involves loss of shielding from impact damage due to tornado-generated missiles. Cask containment loss due to a tornado is not postulated except long-term loss of heat transfer due to blockage of air inlets as discussed in following paragraphs.

There is no fully immersing flood that might move or tip-over the cask postulated for the ANO site. The Maximum Probable Flood blocks the air inlets of the Holtec casks above site Elevation 354 feet.

The VSC-24 storage cask drop accident is a cask drop of 5 feet onto an essentially unyielding surface. The Holtec storage cask drop accident is a cask drop of 11 inches onto an essentially unyielding surface. Any similar drop or tipover of a loaded canister while being transported in a site transfer cask can also potentially affect a confinement boundary.

The full blockage of air inlets event is a postulated blockage of the airflow inlets for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the VSC-24 casks and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (or 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with the difference between the average air outlet temperature and the ISFSI ambient temperature equal to or greater than 126 0 F) for the Holtec casks. The cask has four air inlets and the classification is not based on a loss of confinement boundary, but the condition could lead to the degradation of the fuel during storage or posing an operational safety problem with respect to its removal from storage.

A fire inside the ISFSI fence or explosion that generates missiles that enter the ISFSI area could lead to the degradation of the fuel during storage or pose an operational safety problem with respect to its removal from storage.

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/WORK PLAN TITLE: PAGE: 81 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 EVENTS RELATED TO ISFSI E-HU2 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Confirmed security event with potential loss of level of safety of the ISFSI Operating Mode Applicability:

Not applicable Emergency Action Level:

1. Security event as determined from the ANO Safeguards Contingency Plan and reported by the ANO Security Shift Commander.

Basis:

This EAL is based on ANO Security Plans. Security events which do not represent a potential degradation in the level of safety of the ISFSI are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72.

Reference is made to ANO Security Shift Commander because these individuals are the designated personnel qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the Security Plan.

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MWORK PLAN TITLE: PAGE: 82 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION FU1 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

ANY loss or ANY potential loss of containment Operating Mode Applicability:

Power Operations (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level:

Comparison of conditions/values with those listed in fission product barrier matrix indicates:

Loss or potential loss of containment.

Containment Barrier EALs: CNB1 OR CNB2 OR CNB3 OR CNB4 OR CNB5 OR CNB6 OR CNB7

. Basis:

The fuel cladding and the reactor coolant system are weighted more heavily than the containment barrier.

Loss of the containment would be a potential degradation in the level of plant safety.

PROC.MWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 83 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION FAl ALERT Initiating Condition:

ANY loss or ANY potential loss of EITHER fuel clad or RCS Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level: 1 OR 2 Comparison of conditions/values with those listed in fission product barrier matrix indicates:

1. Loss or potential loss of fuel clad.

OR

2. Loss or potential loss of RCS.

Fuel Clad Barrier EALs: FCB1 OR FCB2 OR FCB3 OR FCB4 OR FCB5 OR FCB6 OR FCB7 OR RCS Barrier EALs: RCB1 OR RCB2 OR RCB3 OR RCB4 OR RCB5 Basis:

The fuel cladding and the reactor coolant system are weighted more heavily than the containment barrier.

Loss of either the fuel cladding or the reactor coolant system would be a substantial degradation in the level of plant safety.

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/WORK PLAN TITLE: PAGE: 84 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 .

Attachment 3 FISSION PRODUCT.BARRIER DEGRADATION FS1 SITE AREA EMERGENCY Initiating Condition:

Loss or potential loss of ANY two barriers Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level: ANY 2 of the 3 Comparison of conditions/values with those listed in fission product barrier Matrix indicates ANY 2 of the following:

Loss or Potential Loss of the fuel clad.

Loss or Potential Loss of the RCS.

Loss or Potential Loss of the containment.

Fuel Clad Barrier EALs: FCB1 OR FCB2 OR FCB3 OR FCB4 OR FCB5 OR FCB6 OR FCB7 RCS Barrier EALs: RCB1 OR RCB2 OR RCB3 OR RCB4 OR RCB5 Containment Barrier EALs: CNB1 OR CNB2 OR CNB3 OR CNB4 OR CNB5 OR CNB6 OR CNB7 Basis:

Loss of 2 fission product barriers would be a major failure of plant systems needed for protection of the public.

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 85 of 156 1903.010 EMERGENCY ACTION q LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION FG1 GENERAL EMERGENCY Initiating Condition:

Loss of ANY two barriers AND loss or potential loss of third barrier Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level: 1 AND 2 Comparison of conditions/values with those listed in fission product barrier matrix indicates:

1. Loss of 2 fission product barriers.

AND

2. Loss or potential loss of third.

Fuel Clad Barrier EALs: FCB1 OR FCB2 OR FCB3 OR FCB4 OR FCB5 OR FCB6 OR FCB7 -

RCS Barrier EALs: RCB1 OR RCB2 OR RCB3 OR RCB4 OR RCB5 Containment Barrier EALs: CNB1 OR CNB2 OR CNB3 OR CNB4 OR CNB5 OR CNB6 OR CNB7 Basis:

Conditions/events causing the loss of 2 Fission Product Barriers with the loss or potential loss of the third could reasonably be expected to cause a release beyond the immediate site area exceeding EPA Protective Action Guidelines.

PROC.MORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 86 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX.XX.0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION FUEL CLAD BARRIER EALs: FCB1 OR FCB2 OR FCB3 OR FCB4 OR FCB5 OR FCB6 OR FCB7 The fuel clad barrier is the zircalloy tubes that contain the fuel pellets.

1. Safety Function Status/Functional Recovery (FCBI)

Loss: None Potential Loss:

ANO-1: HPI Cooling is required to be initiated.

ANO-2: Inadequate RCS heat removal via SGs leads to initiation of once-through cooling.

Basis:

There is no Loss EAL for this item.

The potential loss EAL for this item is a significant challenge to the ability to remove heat from the RCS, and therefore represents a potential challenge to both the fuel clad barrier and the RCS barrier.

ANO-2 EOP 2202.009, "Functional Recovery", contains success paths for RCS heat removal by the steam generators with and without SIAS (HR-2 and HR-1, respectively). An effective SG heatsink is defined as a SG having enough secondary inventory with steaming capability such that core decay heat can be removed without uncontrolled RCS temperature rise. Upon determination that RCS heat removal via unisolated or intact SGs is NOT adequate, the operator is directed to success path HR-3, Once Through Cooling. The criteria used for adequate RCS heat removal via unisolated or intact SGs is based on at least ONE SG with level greater than 70 inches (120 inches under harsh conditions) AND RCS Tc NOT rising in an uncontrolled manner. ANO-2 EOP 2202.006, "Loss of Feedwater", uses similar criteria for initiating once-through cooling. The SGs are the preferred means of core heat removal, and once-through cooling is the method of last resort for core cooling.

Note that this criterion is also considered a challenge to the RCS barrier in the RCS potential loss EAL, RCB1. Therefore, this EAL condition represents a potential loss of both the fuel clad and the RCS barriers, and represents a Site Area Emergency per FS1.

Similarly, the ANO-1 EOP 1202.004, "Overheating", attempts to recover from a challenge to the heat sink, including CET temperatures rising above 610'F AND all MFW/EFW is lost during a loss of adequate SCM, or loss of all feedwater (MRF and EFW) following a reactor trip. Heat removal via the SGs is the preferred means for cooling the core. Upon failure of actions to correct

PROCJWORK PLAN NO. PROCEDUREJWORK PLAN TITLE: PAGE: 87 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-O Attachment 3 FISSION PRODUCT BARRIER DEGRADATION (FCB1) overheating as evidenced by the ERV opening, RCS pressure greater than or equal to 2450 psig, or RCS pressure approaching the NDTT Limit of EOP Figure 3, or secondary feed is NOT expected to become available, the operator is directed to initiate HPI cooling. HPI cooling involves adding relatively cold water to the RCS with the HPI system while removing relatively hot water through the ERV, and can result in releasing large quantities of RCS to the reactor building. Additionally, HPI cooling will probably not initially match the decay heat rate.

Note that this criterion is also considered a challenge to the RCS barrier in the RCS potential loss EAL, RCB1. Therefore, this EAL condition represents a potential loss of both the fuel clad and the RCS barriers, and represents a Site Area Emergency per FS1.

Reference Documents

1. ANO-1 EOP 1202.004, "OVERHEATING"

2. ANO-2 EOP 2202.006, "LOSS OF FEEDWATER"

3. ANO-2 EOP 2202.009, "FUNCTIONAL RECOVERY"

4. BWOG EOP Technical Bases Document, Vol. 3, Chapter III.C

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TiTLE: PAGE: 88 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION FUEL CLAD BARRIER EALs: FCB1 OR FCB2 OR FCB3 OR FCB4 OR FCB5 OR FCB6 OR FCB7 The fuel clad barrier is the zircalloy tubes that contain the fuel pellets.

2. Primary Coolant Activity Level (FCB2)

Loss:

Greater than 300 iiCi/gm dose equivalent I-131 activity by Chemistry sample.

OR ANO-1: Radiation levels at SA-229 indicate greater than 1000 mR/hr.

OR ANO-2: Radiation levels at 2TCD-19 indicate greater than 1000 mR/hr.

Potential Loss: None Basis:

An RCS concentration of 300 pCi/gm dose equivalent I-131 has been determined to correspond to approximately 2.9% failed clad for ANO-1, and 2.1% clad damage for ANO-2, which is consistent with the NUMARC EAL Task Force Assessment that this level corresponds to less than 5% clad damage. This amount of radioactivity is well above that expected for iodine spikes and thus indicates significant clad damage and thus the fuel clad barrier is considered lost.

A reading of greater than 1000 mR/hr within at one foot from the RCS sample lines (SA-229 for ANO-1, 2TCD-19 for ANO-2) has been determined to correspond to fuel clad failure of approximately 2-5%, and thus the fuel clad barrier is considered lost. This reading is well above that, expected for iodine spikes and thus indicates significant clad damage and thus the fuel clad barrier is considered lost.

There is no equivalent potential loss EAL for this item.

Reference Documents

1. ANO Calculation 03-E-0002-01, Radiation Monitor EAL Setpoints for Fission Product Barrier Degradation

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IWORK PLAN TITLE: PAGE: 89 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION FUEL CLAD BARRIER EALs: FCB1 OR FCB2 OR FCB3 OR FCB4 OR FCB5 OR FCB6 OR FCB7 The fuel clad barrier is the zircalloy tubes that contain the fuel pellets.

3. Core Exit Thermocouple Readings (FCB3)

Loss: ANO-1: Greater than or equal to 1200*F CET temperature OR Significant ICC exists as evidenced by CETs indicating superheated conditions ANO-2: Greater than or equal to 1200 0 F average CET Temperature Potential loss: ANO-1: ICC exists as evidenced by CETs indicating superheated conditions ANO-2: Average CETs indicate superheat for current RCS pressure.

Basis:

The loss EAL reading corresponds to significant superheating of the coolant.

The loss EAL of greater than or equal to 1200'F for ANO-2 is consistent with the generic value and is also consistent with recommendations from CE in reference document #5'. The elevated temperature corresponds to significant superheating of the coolant and is indicative of a loss of the fuel clad barrier. Figure 5-2 of reference document #5 is the bases for Figure 1-2 of reference document #4, used to estimate core damage using core exit thermocouples for either unit, and indicates that clad rupture due to high temperature is not expected for CET temperature readings of less than 1200'F.

For ANO-1, the loss EAL is consistent with the treatment of inadequate core cooling (ICC) in the EOPs, which is based on a pressure-temperature curve. The basis for Region 3 of this curve from the BWOG EOP Technical Basis Document states, "If the RCS P-T reaches Region Three, then cladding temperature in the high power regions of the core may be 14000 F or higher." This is consistent with the intent of the 1200*F CET reading recommendation, as CET temperature will be lower than fuel clad temperature..

The potential loss EAL corresponds to a loss of subcooling. For ANO-2, there is a Functional Recovery EOP (2202.009), and the core and RCS heat removal acceptance criteria for safety function status checks include determination of RCS superheated.

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. ~CHANGE: XXX-XX-0-.

Attachment 3 FISSION PRODUCT BARRIER DEGRADATION (FCB3)

For ANO-1, the RCS P-T in Region 2 (CET temperatures above saturation for indicated pressure). of the EOP Figure 4 corresponds to a loss of subcooling.

This is consistent with EOP 1202.005, "Inadequate Core Cooling".

Note that the loss or potential loss EAL for this category will occur after a loss of adequate sub-cooling margin, which represents a loss of the RCS barrier in EAL RCB2, and therefore represents the loss of two barriers, resulting in a Site Area.Emergency per FS1. Any loss or potential loss of the containment barrier at that point would escalate to a General Emergency..

Reference Documents

1. ANO-1 EOP 1202.005, "Inadequate Core Cooling"

2. ANO-1 EOP 1202.013, EOP Figures

3. ANO-2 OP 2202.009, "Functional Recovery"

4. ANO Procedure OP 1302.022, "Core Damage Assessment"

5. CE-NPSD-241, Development of the Comprehensive Procedure Guideline for Core Damage Assessment, Task 467

6. BWOG EOP Technical Bases Document, Vol. 3, Chapter III.F

PROCJWORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 91 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION FUEL CLAD BARRIER EALs: FCB1 OR FCB2 OR FCB3 OR FCB4 OR FCB5 OR FCB6 OR FCB7 The fuel clad barrier is the zircalloy tubes that contain the fuel pellets.

4. Reactor Vessel Water Level (FCB4)

Loss: None Potential Loss: If CET indication is unavailable AND all RCPs are secured, indication of core uncovery:

ANO-1: All RVLMS sensors indicate DRY following lowering trend ANO-2: RVLMS LVL6 indicates DRY following lowering trend Basis:

The Reactor Vessel Level Monitoring Systems at ANO do not provide positive indication of core uncovery. The above core level indication provided is used to monitor the approach to and recovery from ICC conditions, but the CETs are used to identify core uncovery, and are the only positive indication of core uncovery. Consistent with this approach, RVLMS is used as an indication of potential core uncovery only if CET indication is unavailable.

Per reference document #1, the reactor vessel level indicators installed in ANO-1 extend from the top of the reactor vessel to the fuel alignment plate, and information in reference document #2 indicates that the lowest sensor is greater than 2 feet above the top of active fuel. If any of the 4 RCPs are running, flow induced turbulence produced by the pumps renders the reactor vessel level indicator readings invalid.

Per reference document #3, only the reactor vessel level indicators above the core are considered part of the ICC monitoring system. Per reference document

4, the lowest sensor above the core, RVLMS LVL 6 on the ICC monitoring panel 2C388, is 47 inches above the top of the core. If any-of the 4 RCPs are running, flow induced turbulence produced by the pumps renders the reactor vessel level indicator readings invalid.

For either unit then, should CET indication be unavailable and reactor vessel level indication be unavailable due to RCP operation or any other cause, a degraded ability to monitor the barrier would exist.

Reference Documents

1. ULD-1-SYS-24, ANO-1 Inadequate Core Cooling

2. Calculation 84-EQ-0080-02, Loop Error Analysis for Reactor Vessel Level Monitoring System

3. ULD-2-SYS-24, ANO-2 Inadequate Core Cooling

4. Calculation 90-E-0116-01, ANO-2 EOP Setpoint Document, Setpoint R.3

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 92 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION FUEL CLAD BARRIER EALs: FCB1 OR FCB2 OR FCB3 OR FCB4 OR FCB5 OR FCB6 OR FCB7 The fuel clad barrier is the zircalloy tubes that contain the fuel pellets.

5. Containment Radiation Monitoring (FCB5)

Loss: Containment high range rad monitor reading greater than 1000 R/hr Potential Loss: NONE Basis:

The 1000 R/hr reading on the containment high range radiation monitors (RE-8060 or RE-8061 for ANO-1, 2RE-8925-1 or 2RE-8925-2 for ANO-2) is a value which indicates the release of reactor coolant, with elevated activity indicative of fuel damage, into the containment. The reading was calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with approximately 2-5% cladding failure into the containment atmosphere. Reactor coolant concentrations of this magnitude are several times larger than that expected for iodine spikes and are therefore indicative of fuel damage. This value is higher than that specified for RCS barrier loss EAL RCB4. Therefore, this EAL condition represents a potential loss of both the fuel clad and the RCS barriers, and represents a Site Area Emergency per FS1.

There is no potential loss EAL associated with this item.

Reference Documents

1. NUREG 1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents

2. ANO Calculation 03-E-0002-01, Radiation Monitor EAL Setpoints for.

Fission Product Barrier Degradation

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 93 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION FUEL CLAD BARRIER EALs: FCB1 OR FCB2 OR FCB3 OR FCB4 OR FCB5 OR FCB6 OR FCB7 The fuel clad barrier is the zircalloy tubes that contain the fuel pellets.

6. Core Damage Assessment (FCB6)

Loss: At least 5% fuel clad damage as determined from core damage assessment Potential Loss: NONE Basis:

This level is consistent with other fuel clad barrier loss EALs indicative of significant fuel clad damage, but uses core damage assessment evaluations by Technical Support personnel. The fuel clad-barrier is considered lost.

If this determination is made from the high range containment radiation monitor readings, or if accompanied by other indications of a loss or potential loss of the RCS barrier, this EAL condition represents a Site Area Emergency per FS1.

There is no potential loss EAL associated with this item.

Reference Documents

1. ANO Procedure OP-1302.022, "Core Damage Assessment"

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JWORK PLAN TITLE: PAGE: 94 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION FUEL CLAD BARRIER EALs: FCB1 OR FCB2 OR FCB3 OR FCB4 OR FCB5 OR FCB6 OR FCB7 The Fuel Clad Barrier is the zircalloy tubes that contain the fuel pellets.

7. SM/TSC Director/EOF Director Judgment (FCB7)

Any condition in the opinion of the SM/TSC Director/EOF Director that indicates loss or potential loss of the fuel clad barrier based on:

Imminent barrier degradation (within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) due to degraded safety system performance

Degraded ability to monitor barrier Basis:

This EAL addresses any other factors that are to be used by the SM/TSC Director/EOF Director in determining whether the fuel clad barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL as a factor in the SM/TSC Director/EOF Director judgment that the barrier may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss or All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional information.)

PROCAWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 95 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION RCS BARRIER EALs: RCB1 OR RCB2 OR RCB3 OR RCB4 OR RCB5 The RCS barrier is the reactor coolant system pressure boundary and includes the reactor vessel and all reactor coolant system piping up to the isolation valves.

1. Safety Function Status/Functional Recovery (RCB1)

Loss: None Potential Loss ANO-1: HPI cooling is required to be initiated OR RCS Pressure greater than 2450 PSIG AND not lowering ANO-2: Inadequate RCS heat removal via SGs leads to initiation of once-through cooling OR RCS Pressure greater than 2465 PSIA AND not lowering Basis:

There is no loss EAL associated with this item.

The first potential loss EAL for this item is a significant challenge to the ability to remove heat from the RCS, and therefore represents a potential challenge to both the fuel clad and the RCS barriers.

ANO-2 EOP 2202.009, "Functional Recovery", contains success paths for RCS heat removal by the steam generators with and without SIAS (HR-2 and HR-1, respectively). An effective SG heatsink is defined as a SG having enough secondary inventory with steaming capability such that core decay heat can be removed without uncontrolled RCS temperature increase. Upon determination that RCS heat removal via unisolated or intact SGs is NOT adequate, the operator is directed to success path HR-3, Once Through Cooling. The criteria used for adequate RCS heat removal via unisolated or intact SGs is based on at least ONE SG with level greater than 70 inches (120 inches under harsh conditions) AND RCS Tc NOT rising in an uncontrolled manner. ANO-2 EOP 2202.006, "Loss of Feedwater", uses similar criteria for initiating once-through cooling. The SGs are the preferred means of core heat removal, and once-through cooling is the method of last resort for core cooling.

PROC.IWORK PLAN NO. PROCEDUREJNORK PLAN TITLE: PAGE: 96 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION (RCB1)

Note that this criteria is also considered a challenge to the fuel clad barrier in the fuel clad barrier potential loss EAL, FCB1. Therefore this first potential loss EAL condition represents a potential loss of both the' fuel clad and the RCS barriers, and represents a Site Area Emergency per FS1.

Similarly, the ANO-1 EOP 1202.004, "Overheating", attempts to recover from a challenge to the heat sink, including CET temperatures rising above 610'F AND all MFW/EFW is lost during a loss of adequate SCM, or loss of all feedwater (MFW and EFW) following a reactor trip. Heat removal via the SGs is the preferred means for cooling the core. Upon failure of actions to correct overheating as evidenced by the ERV opening, RCS pressure greater than or equal to 2450 psig, RCS pressure approaching the NDTT Limit of EOP Figure 3, or secondary feed is NOT expected to become available, the operator is directed to initiate HPI cooling. HPI cooling involves adding relatively cold water to the RCS with the HPI system while removing relatively hot water through the ERV, and can result in releasing large quantities of RCS to the reactor building. Additionally, HPI cooling will probably not initially match the decay heat rate.

Note that this criteria is also considered a challenge to the fuel clad barrier in the fuel clad potential loss EAL FCB1. Therefore this EAL condition represents a potential loss of both the fuel clad and the RCS barriers, and represents a Site Area Emergency per FS1.

Historically, ANO-l and ANO-2 have regarded that RCS pressure greater than 2450 PSIG *(Unit 1) and greater than 2465 PSIA (Unit 2) and NOT lowering represents a challenge to RCS integrity, in that it represents a possible uncontrolled overpressurization of the RCS. For ANO-1, the combination of the ERV setpoint, the pressurizer code safety setpoints, the RCS high pressure trip, and the DSS high pressure trip, in conjunction with recovery actions are all expected to be able to lower RCS pressure below 2450 PSIG. For ANO-2, the pressurizer code safety setpoints, the RCS high pressure trip, and the DSS high pressure trip, in conjunction with recovery actions are all expected to be able to lower RCS pressure below 2465 PSIA.

Reference Documents

1. ANO-1 EOP 1202.004, "Overheating"

2. ANO-2 EOP 2202.006, "Loss Of Feedwater"

3. ANO-2 EOP 2202.009, "Functional Recovery"

4. CEN-152, Emergency Operating Procedure Guidelines

5. Calculation 90-E-0116-01, ANO-2 EOP Setpoint Document

6. BWOG EOP Technical Bases Document, Vol. 3, Chapter III.C

PROCJWORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 97 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION RCS BARRIER EALs: RCB1 OR RCB2 OR RCB3 OR RCB4 OR RCB5 The RCS barrier is the reactor coolant system pressure boundary and includes the reactor vessel and all reactor coolant system piping up to the isolation valves.

2. RCS Leak Rate (RCB2)

Loss: RCS leakage greater than available makeup capacity as indicated by:

ANO-1: Loss of adequate subcooling margin ANO-2: RCS subcooling (MTS) can NOT be maintained at least 30'F Potential Loss:

ANO-1: RCS leakage exceeding Normal Makeup Capacity (50 gpm)

ANO-2: RCS leakage exceeding the capacity of one charging pump in the normal charging mode (44 gpm)

Basis:

The loss EAL addresses conditions where leakage from the RCS is greater than available inventory control capacity such that a loss of subcooling has occurred. The loss of subcooling is the fundamental indication that the inventory control systems are inadequate in maintaining RCS pressure and inventory against the mass loss through the leak.

The potential loss EAL is based on the inability to maintain normal liquid inventory within the reactor coolant system (RCS) by normal operation of the Makeup and Purification System (Unit 1) or the Chemical and Volume Control System (Unit 2).

For ANO-1 this is based on indications that leakage is greater than normal makeup capacity. The operator could not batch in water and boric acid to the makeup system fast enough to maintain the makeup tank level during a 50 gpm RCS leak. It is not necessary to perform a detailed assessment of the RCS leakrate to implement this EAL. Any event or condition which, in the judgment of the SM/TSC Director/EOF Director, could result in RCS leakage in excess of ANO-1 normal makeup capacity would meet the intent of this EAL; for example:

Need to open the BWST suction for the operating makeup pump due to decreasing makeup tank level

Full or partial HPI is needed to maintain the RCS pressure or pressurizer level

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 98 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION (RCB2)

Two out of three seal stages failed on any RCP

RCS pressure decreasing due to failure of a primary relief valve to reseat For ANO-2, this is considered as the capacity of one charging pump discharging to the charging header (44 gpm). Any event or condition which, in the judgment of the SM/TSC Director/EOF Director, could result in RCS leakage in excess of ANO-2 normal makeup capacity would meet the intent of this EAL; for example:

A second charging pump being required is indicative of a substantial RCS leak

Three out of four seal stages failed on any RCP

RCS pressure decreasing due to failure of a primary relief valve to reseat Reference Documents

1. ANO-1 EOP 1202.013, Figure 1, Saturation and Adequate SCM

2. ANO-1 EOP Setpoint Document, Calculation 90-E-0016-07, Setpoint B.19

3. ANO-2 EOP 2202.009, "Functional Recovery"

4. ANO-2 EOP Setpoint Document, Calculation 90-E-0116-01

5. Unit 2 SAR Table 9.3-14, Charging Pumps Design Data

PROCJWORK PLAN NO. PROCEDUREJWORK PLAN TITLE: PAGE: 99 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION RCS BARRIER EALs: RCB1 OR RCB2 OR RCB3 OR RCB4 OR RCB5 The RCS barrier is the reactor coolant system pressure boundary and includes the reactor vessel and all reactor coolant system piping up to the isolation valves.

3. SG Tube Rupture (RCB3)

Loss: SGTR that results in an ECCS (SI) actuation Potential Loss: NONE Basis:

This EAL is intended to address the full spectrum of steam generator (SG) tube rupture events in conjunction with containment barrier loss EAL CNB3 and fuel clad barrier EALs. The loss EAL addresses RUPTURED SG(s) for which the leakage is large enough to cause actuation of ECCS safety injection. This is consistent to the RCS barrier potential loss EAL RCB2. By itself, this EAL will result in the declaration of an Alert. However, if the SG is also FAULTED (i.e., two barriers failed), the declaration escalates to a Site Area Emergency per containment barrier loss EAL CNB3.

There is no potential loss EAL.

PROCJWORK PLAN NO. PROCEDUREMWORK PLANnTLE: PAGE: 100 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION RCS BARRIER EALs: RCB1 OR RCB2 OR RCB3 OR RCB4 OR RCB5 The RCS barrier is the reactor coolant system pressure boundary and includes the reactor vessel and all reactor coolant system piping up to the isolation valves.

4. Containment Radiation Monitoring (RCB4)

Loss: Containment rad monitor reading greater than .100 R/hr Potential Loss: NONE Basis:

The 100 R/hr reading on the containment high range radiation monitors (RE-8060 or RE-8061 for ANO-1, 2RE-8925-l or 2RE-8925-2 for ANO-2) is a value which indicates the release of reactor coolant to the containment. This reading was derived assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory associated with an RCS concentration of 60 pCi/gm dose equivalent I-131 into the containment atmosphere. This reading is an order of magnitude lower than that specified for fuel clad barrier EAL FCB5. Thus, this EAL would be indicative of an RCS leak only. If the radiation monitor reading increased to that specified by fuel clad barrier EAL FCB5, fuel damage would also be indicated.

During the initial fifteen minutes after a thermal event inside containment, the high range radiation monitor readings are considered invalid due to possibility of a transient thermally-induced current.

There is no potential loss EAL associated with this item.

Reference Documents

1. ANO Calculation 03-E-0002-01, Radiation Monitor EAL Setpoints for Fission Product Barrier Degradation

PROC.MVORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 101 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-.

Attachment 3 FISSION PRODUCT BARRIER DEGRADATION RCS BARRIER EALs: RCB1 OR RCB2 OR RCB3 OR RCB4 OR RCB5 The RCS barrier is the reactor coolant system pressure boundary and includes the reactor vessel and all reactor coolant system piping up to the isolation valves.

5. SM/TSC Director/EOF Director Judgment (RCB5)

Any condition in the opinion of the SM/TSC Director/EOF Director that indicates loss or potential loss of the RCS barrier based on:

Imminent barrier degradation (within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) due to degraded safety system performance

Degraded ability to monitor barrier Basis:

This EAL addresses any other factors that are to be used by the SM/TSC Director/EOF Director in determining whether the RCS barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL as a factor in SM/TSC Director/EOF Director judgment that the barrier may be considered lost or potentially lost. '(See also IC SG1, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power", for additional information.)

PROCJWORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 102 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT BARRIER EALs: CNB1 OR CNB2 OR CNB3 OR CNB4 OR CNB5 OR CNB6 OR CNB7 -

The containment barrier includes the containment building, its connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.

1. Containment Pressure (CNB1)

Loss:

Rapid unexplained containment pressure loss following initial rise OR Containment parameters not consistent with expected event response Potential Loss:

ANO-1: 73.7 PSIA (59 PSIG) and rising ANO-2: 73.7 PSIA and rising OR Containment Hydrogen Concentration greater than 4.0%

OR Containment Pressure greater than containment spray actuation setpoint with less than one full train of spray operating ANO-1 44.7 PSIA (30 PSIG)

ANO-2 23.3 PSIA Basis:

Rapid unexplained loss of pressure (i.e., not attributable to containment spray or condensation effects) following an initial pressure rise indicates a loss of containment integrity. Containment pressure and sump levels should rise as a result of the mass and energy release into containment from a LOCA.

Thus, sump level or pressure or humidity (ANO-2) not rising indicates containment bypass and a loss of containment integrity. The containment pressure setpoint for potential loss of containment is based on the containment design pressure. The hydrogen concentration of 4% has been recognized by the NRC staff as a well-established lower flammability limit in air or steam-air atmospheres that is adequately conservative for protecting against an H2 explosion. Hydrogen control systems at ANO are

PROCJ\WORK PLAN NO. PROCEDUREJWORK PLAN WTLE: PAGE: 103 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: xxx-xx-O Attachment 3 FISSION PRODUCT BARRIER DEGRADATION (CNB1) designed and operated as to maintain the containment hydrogen concentration below this level, so that indications of hydrogen concentrations above this are considered a potential challenge to the containment integrity.

Conditions leading to these indications result from RCS barrier and/or fuel clad barrier loss. Thus, this EAL is primarily a discriminator between Site Area Emergency and General Emergency representing a potential loss of the third barrier.

The second potential loss EAL based on containment pressure represents a potential loss of containment in that the containment heat removal/depressurization system (containment sprays, but not including containment venting strategies) are either lost or performing in a degraded manner, as indicated by containment pressure greater than the setpoint at which the equipment was supposed to have actuated.

Reference Documents

1. ANO-1 OP-1105.003, "Engineering Safeguards Actuation System"

2. ANO-1 SAR Sections 1.4.43, 5.2.1.2.1, 14.2.2.5.5.1 (reactor building design pressure)

3. ANO-1 SAR Section 6.6 Post-Loss of Coolant Accident Hydrogen Control

4. ANO-1 TS Table 3.3.5-1

5. ANO-2 SAR Section 6.2.5 Combustible Gas Control In Containment

6. ANO-2 SAR Section 3.8.1.3.1.D (Containment Design Pressure)

7. ANO-2 TS Table 3.3-4

8. Regulatory Guide 1.7, Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident, Rev. 2 1978

PROCJNORK PLAN NO. PROCEDUREMVORK PLAN TITLE: PAGE: 104 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT BARRIER EALs: CNB1 OR CNB2 OR CNB3 OR CNB4 OR CNB5 OR CNB6 OR CNB7 The Containment Barrier includes the containment building, its connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.

2. Core Exit Thermocouples (CNB2)

Loss: None Potential Loss:

ANO-1: Significant ICC exists as evidenced by CETs indicating superheated conditions and restoration procedures not effective within 15 minutes ANO-2: CETs greater than 1200'F AND restoration procedures not effective within 15 minutes Basis:

In this EAL, the function restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety functions. The procedure is considered effective if the temperature is decreasing.

Severe accident analyses (e.g., NUREG-1150) have concluded that function restoration procedures can arrest core degradation within the reactor vessel in a significant fraction of the core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this, it is appropriate to provide a reasonable period to allow function restoration procedures to arrest the core melt sequence. Whether or not the procedures will be effective should be apparent withini15 minutes. The SM/TSC Director/EOF Director should make the declaration as soon as it is determined that the procedures have been, or will be ineffective.

The conditions in this potential loss EAL represent an imminent core melt sequence which, if not corrected, could lead to vessel failure and a higher potential for containment failure. In conjunction with the core cooling and heat sink criteria in the fuel and RCS barrier columns, this EAL would result in the declaration of a General Emergency (loss of two barriers and the potential loss of a third). If the function restoration procedures are ineffective, there is no success path.

PROCJWORK PLAN NO. PROCEDURENWORK PLAN TITLE: PAGE: 105 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT BARRIER EALs: CNB1 OR CNB2 OR CNB3 OR CNB4 OR CNB5 OR CNB6 OR CNB7)

The containment barrier includes the containment building, its connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.

3. SG Secondary Side Release With Primary to Secondary Leakage (CNB3)

Loss: Primary-to-secondary leakrate greater than 10 gpm with nonisolable steam release from affected SG to the environment Potential Loss: NONE Basis:

This loss EAL recognizes that SG tube leakage can represent a bypass of the containment barrier as well as a loss of the RCS barrier. Secondary side release paths to environment include atmospheric relief valves and main steam line safety valves, as well as discharges direct to the environment from an unisolable secondary or steam line break. The threshold for establishing the nonisolable secondary side release is intended to be a prolonged release of radioactivity from the RUPTURED steam generator directly to the environment.

This could be expected to occur when the main condenser is unavailable to accept the contaminated steam (i.e., SGTR with concurrent loss of offsite power and the RUPTURED steam generator is required for plant cooldown or has a stuck open relief valve). If the main condenser is available, there may be releases via air ejectors, gland seal exhausters, and other similar controlled, and often monitored, pathways. These pathways do not meet the intent of a nonisolable release path to the environment. These minor releases are assessed using Abnormal Rad Levels/Radiological Effluent ICs.

For smaller breaks, not exceeding the Normal Makeup Capacity for ANO-l or the capacity of one charging pump in the normal charging lineup for ANO-2, but exceeding 10 gpm, this EAL results in an Unusual Event.

For breaks that exceed the Normal Makeup Capacity for ANO-1 or the capacity of onecharging pump in the normal charging lineup for ANO-2 or result in ECCS actuation, RCS barrier EALs RCB2 or RCB3 would result in an Alert if the ruptured SG is isolated. If the SG remains unisolated, this EAL will be a discriminator for Site Area Emergencies. Escalation to General Emergency would be based on Loss or Potential Loss of the fuel clad barrier.

There is no equivalent potential loss EAL for this item.

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 106 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION

. ~CHANGE: XXXK-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT BARRIER EALs: CNB1 OR CNB2 OR CNB3 OR CNB4 OR CNB5 OR CNB6 OR CNB7 The containment barrier includes the containment building, its connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.

4. Containment Isolation Valve Status after Containment Isolation (CNB4)

Loss: Unisolable breach of containment with a direct release path to the environment following containment isolation actuation Potential Loss: NONE Basis:

This EAL is intended to address incomplete containment isolation that allows direct release to the environment. It represents a loss of the containment barrier. A breach of containment has also occurred if an inboard and outboard pair of isolation valves fails to close on an automatic actuation signal or from a manual action in the control room and opens a release path to the environment. This EAL is not intended to prohibit overriding containment isolation valves when directed by plant procedures. A manually overridden containment isolation valve is considered isolable until proven otherwise.

The breach is not isolable from the Control Room if an attempt for isolation from the Control Room has been made and was unsuccessful. An attempt for isolation should be made prior to the accident classification. If isolable upon identification then this Initiating Condition is not applicable.

There is no potential loss EAL associated with this item.

PROCNJORK PLAN NO. PROCEDURE!NORK PLAN TITLE: PAGE: 107 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT BARRIER EALs: CNB1 OR CNB2 OR CNB3 OR CNB4 OR CNB5 OR CNB6 OR CNB7 The containment barrier includes the containment building, its connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.

5. Significant Radioactive Inventory in Containment (CNB5)

Loss: None Potential Loss: Containment high range Rad Monitor reading greater than 4000 R/hr Basis:

The 4000 R/hr reading on the containment high range radiation monitors (RE-8060 or RE-8061 for ANO-1, 2RE-8925-1 or 2RE-8925-2 for ANO-2) is a value which indicates significant fuel damage well in excess of the EALs associated with both loss of fuel clad and loss of RCS barriers. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant.

Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%.

There is no loss EAL associated with this item.

Reference Documents:

1. ANO Calculation 03-E-0002-01, Radiation Monitor EAL Setpoints for Fission Product Barrier Degradation

2. NUREG 1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 108 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT BARRIER EALs: CNB1 OR CNB2 OR CNB3 OR CNB4 OR CNB5 OR CNB6 OR CNB7 The containment barrier includes the containment building, its connections up.

to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.

6. Core Damage Assessment (CNB6)

Loss: None Potential Loss: At least 20% fuel clad failure as determined from core damage assessment Basis:

Twenty percent fuel cladding failure is a value which indicates significant fuel damage well in excess of the EALs associated with both loss of fuel clad and loss of RCS barriers. A major release of radioactivity requiring offsite protective actions from core damage is not possible unless a major failure of fuel cladding allows radioactive material to be released from the core into the reactor coolant.

Regardless of whether containment is challenged, this amount of activity in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted. NUREG-1228, "Source Estimations During Incident Response to Severe Nuclear Power Plant Accidents," indicates that such conditions do not exist when the amount of clad damage is less than 20%.

This EAL is consistent with the intent of EAL CNB5, but uses core damage assessment evaluations by Technical Support personnel.

There is no loss EAL associated with this item.

Reference Documents

1. NUREG 1228, Source Term Estimation During Incident Response to Severe Nuclear Power Plant Accidents

2. ANO Procedure OP-1302.022, "Core Damage Assessment"

PROCJWORK PLAN NO. PROCEDUREJWORK PLAN TITLE: PAGE: 109 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-O Attachment 3 FISSION PRODUCT BARRIER DEGRADATION CONTAINMENT BARRIER EALs: CNB1 OR CNB2 OR CNB3 OR CNB4 OR CNB5 OR CNB6 OR CNB7 The containment barrier includes the containment building, its connections up to and including the outermost containment isolation valves. This barrier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation valve.

7. Emergency Director Judgment (CNB7)

Any condition in the opinion of the SM/TSC Director/EOF Director that indicates loss or potential loss of the containment barrier based on:

Imminent barrier degradation (within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) due to degraded safety system performance

Degraded ability to monitor barrier Basis:

This EAL addresses any other factors that are to be used by the SM/TSC Director/EOF Director in determining whether the containment barrier is lost or potentially lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL as a factor in SM/TSC Director/EOF Director judgment that the barrier may be considered lost or potentially lost. (See also IC SG1, "Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite.

AC Power", for additional information.)

PROCJWORK PLAN NO. PROCEDUREJWORK PLAN TITLE: PAGE: 110 of 156 1903.010 EMERGENCY ACTION LEVELCLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety HUI NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Confirmed security event which indicates a potential degradation in the level of safety of the plant Operating Mode Applicability:

All Emergency Action Level: 1 OR 2

1. Security events as determined from the Safeguards Contingency Plan and reported by the Security Shift Commander.

2. A credible security threat notification.

Basis:

The Security Shift Commander is the designated individual on-site qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the plant Safeguards Contingency Plan.

EAL #1 is based on the Site Security Plan. Security events which do not represent a potential degradation in the level of safety of the plant, are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. Examples of security events that indicate potential degradation in the level of safety of the plant are provided below for consideration.

Consideration should be given to the following types of events when evaluating an event against the criteria of the Security Contingency Plan: SABOTAGE, HOSTAGE/EXTORTION, CIVIL DISTURBANCE, and STRIKE ACTION.

INTRUSION into the plant PROTECTED AREA by a HOSTILE FORCE would result in EAL escalation to an ALERT.

The intent of EAL #2 is to ensure that appropriate notifications for the security threat are made in a timely manner. The determination of "credible" is made through use of information found in the Safeguards Contingency Plan.

A higher initial classification could be made based upon the nature and timing of the threat and potential consequences. Consideration shall be given to upgrading the emergency response status and emergency classification in accordance with the Safeguards Contingency Plan and Emergency Plans.

PROCJWORK PLAN NO. PROCEDUREJWORK PLAN TITLE: PAGE: 111 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety HU2 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Other conditions exist which in the judgment of the SM/TSC Director/EOF Director warrant declaration of an NUE Operating Mode Applicability:

All Emergency Action Level:

1. Other conditions exist which in the judgment.of the SM/TSC Director/EOF Director indicate that events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

Basis:

This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the SM/TSC Director/EOF Director to fall under the NUE emergency class.

From a broad perspective, one area that may warrant SM/TSC Director/EOF Director judgment is related to likely or actual breakdown of site-specific event mitigating actions. Examples to consider include inadequate emergency response procedures, transient response either unexpected or not understood, failure or unavailability of emergency-systems during an accident in excess of that assumed in accident analysis, or insufficient availability of equipment and/or support personnel.

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 112 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety HU4 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

FIRE within PROTECTED AREA Boundary not extinguished within 15 minutes of detection Operating Mode Applicability:

All Emergency Action Level:

1. FIRE in Table Hl buildings or areas adjacent to any of Table Hl areas on either unit not extinguished within 15 minutes of Control Room notification or verification of a Control Room alarm:

Table Hi Intake Structure Containment Auxiliary Building Aux Extension Building QCST/RWT/BWST Diesel Fuel Oil Vault Transformer Yard Turbine Building Basis:

The purpose of this IC is to address the magnitude and extent of FIREs that may be potentially significant precursors to damage to safety systems. As used here, Detection is visual observation and report by plant personnel or sensor alarm indication. The 15 minute time period begins with a credible notification that a FIRE is occurring, or indication of a VALID fire detection system alarm.

Verification of a fire detection system alarm includes actions that can be taken within the Control Room to ensure that the alarm is not spurious. A verified alarm is assumed to be an indication of a FIRE unless it is disproved within the 15 minute period by personnel dispatched to the scene. In other words, a personnel report from the scene may be used to disprove a sensor alarm if received within 15 minutes of the alarm, but shall not be required to verify the alarm.

The intent of this 15 minute duration is to size the FIRE and to discriminate against small FIREs that are readily extinguished (e.g., smoldering waste paper basket). Table Hl applies to buildings and areas adjacent (in actual contact with or immediately adjacent) to plant VITAL AREAs or other significant buildings or areas. The intent of this EAL is not to include buildings (i.e., warehouses) or areas that are not adjacent (in actual contact with or immediately adjacent) to plant VITAL AREAs. This IC excludes FIREs within administration buildings, waste-basket FIREs, and other small FIREs of no safety consequence.

Escalation to a higher emergency class is by HA4.

PROC.JWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 113 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety HU5 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Release of toxic or flammable gases deemed detrimental to normal operation of the plant Operating Mode Applicability:

All Emergency Action Level: 1 OR 2

1. -Report or detection of toxic or flammable gases that have or could enter normally occupied areas of the. site in amounts that can affect NORMAL PLANT OPERATIONS.

2. Report by Local, County or State officials for evacuation or sheltering of site personnel based on an offsite event.

Basis:

This IC is based on the existence of uncontrolled releases of toxic or flammable gas that may enter the site boundary and affect normal plant operations.. It is intended that releases of toxic or flammable gases are of sufficient quantity, and the release point of such gases is such that normal plant operations would be affected. This would exclude small or incidental releases, or releases that do not impact structures needed for plant operation. The EALs are intended to not require significant assessment or quantification. The EALs assume an uncontrolled process that has the potential to affect plant operations, or personnel safety.

Escalation of this EAL is via HA5, which involves a quantified release of toxic or flammable gas affecting VITAL AREAs.

PROCJWORK PLAN NO. PROCEDURENVORK PLAN TITLE: PAGE: 114 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION

. CHANGE: XXX-XX-0 .

Attachment 3 Hazards and Other Conditions Affecting Plant Safety HU6 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Natural and destructive phenomena affecting the PROTECTED AREA Operating Mode Applicability:

All Emergency Action Level: 1 OR 2 OR 3 OR 4 OR 5 OR 6 OR 7 OR 8

1. An earthquake is felt and the 0.01g acceleration alarm annunciates indicating an earthquake has occurred.

2. Report by plant personnel of tornado or high winds greater than 67 mph striking within PROTECTED AREA boundary.

3. Vehicle crash into plant structures or systems within PROTECTED AREA boundary.

4. Report by plant personnel of an unanticipated EXPLOSION within PROTECTED AREA boundary resulting in VISIBLE DAMAGE to permanent structure or equipment.

5. Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

6. Uncontrolled flooding in areas of the plant that has the potential to affect safety related equipment needed for the current operating mode.

7. Lake Dardanelle level greater than 345 feet.

8. Lake Dardanelle level less than 335 feet.

Basis:

An NUE would be declared on the basis of the occurrence of an event of sufficient magnitude to be of concern to plant operators. Areas identified in the EALs define the location of the event based on the potential for damage to equipment contained therein. Escalation of the event to an Alert occurs when the magnitude of the event is sufficient to result in damage to equipment contained in the specified location.

EAL #1 is based on damage that may be caused to some portions of the site, but should not affect ability of safety functions to operate. The method of detection is based on instrumentation, validated by a reliable source, or operator assessment. As defined in the EPRI sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, a "felt earthquake" is:

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/WORK PLAN TITLE: PAGE: 11 5 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions-Affecting Plant Safety (HU6)

An earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of control room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated.

EAL #2 is based on the assumption that a tornado striking (touching down) or high winds within the PROTECTED AREA may have potentially damaged plant structures containing functions or systems required for safe shutdown of the plant. The high wind value in EAL #2 is conservatively based on the SAR design basis for Unit 1 of 67 mph. Unit 2 Design basis is 80 mph. If damage is confirmed visually or by other plant indications, the event may be escalated to Alert.

EAL #3 is intended to address crashes of vehicle types large enough to cause significant damage to plant structures containing functions and systems required for safe shutdown of the plant. If the crash is confirmed to affect a plant VITAL AREA, the event may be escalated to Alert.

For EAL #4 only those EXPLOSIONs of sufficient force to damage permanent structures or equipment within the PROTECTED AREA should be considered. No attempt is made in this EAL to assess the actual magnitude of the damage. The occurrence of the EXPLOSION with reports of evidence of damage is sufficient for declaration. The SM/TSC Director/EOF Director also needs to consider any security aspects of the EXPLOSION, if applicable.

EAL #5 is based on main turbine rotating component failures of sufficient magnitude to cause observable damage to the turbine casing or to the seals of the turbine generator. Of major concern is the potential for leakage of combustible fluids (lubricating oils) and gases (hydrogen cooling) to the plant environs. Actual FIREs and flammable gas build up are appropriately classified via HU4 and HU5. Generator seal damage observed after generator purge does not meet the intent of this EAL because it did not impact normal operation of the plant. This EAL is consistent with the definition of a NUE while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment. Escalation of the emergency classification is based on potential damage done by missiles generated by the failure or in conjunction with a steam generator tube rupture. The latter event would be classified by the radiological EALs or fission product barrier EALs.

EAL #6 addresses the effect of flooding caused by internal events such as component failures, equipment misalignment, or outage activity mishaps. The site-specific areas include those areas that contain systems required for safe shutdown of the plant and that are not designed to be wetted or submerged.

Escalation of the emergency classification is based on the damage caused or by access restrictions that prevent necessary plant operations or systems monitoring.

EAL #7 and #8 are based on the levels of Lake Dardanelle at which the site will take specific action to reduce the impact of the lake level on plant safety by initiating plant shutdown.

Reference Documents:

1. OP-1203.025 "Natural Emergencies"

2. OP-2203.008 "Natural Emergencies"

PROCJWORK PLAN NO. PROCEDURENWORK PLAN TITLE: PAGE: 11 6 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety HAI ALERT Initiating Condition:

Confirmed security event within a plant PROTECTED AREA Operating Mode Applicability:

All Emergency Action Level: 1 OR 2

1. INTRUSION into the plant PROTECTED AREA by a HOSTILE FORCE.

2. Other security events as determined from the Safeguards Contingency Plan and .reported by the Security Shift Commander..

Basis:

This class of security events represents an escalated threat to plant safety above that contained in the NUE. A confirmed INTRUSION report is satisfied if physical evidence indicates the presence of a HOSTILE FORCE within the PROTECTED AREA.

Consideration should be given to the following types of events when evaluating an event against the criteria of the Security Contingency Plan: SABOTAGE, HOSTAGE/EXTORTION, and STRIKE ACTION. The Safeguards Contingency Plan identifies numerous events/conditions that constitute a threat/compromise to a Station's security. Only those events that involve actual or potential substantial degradation to the level of safety of the plant need to be considered. The following events would not normally meet this requirement; (e.g., Failure by a Member of the Security Force to carry out an assigned/required duty, internal disturbances, loss/compromise of safeguards materials or strike actions).

INTRUSION into a VITAL AREA by a HOSTILE FORCE will escalate this event to a Site Area Emergency.

The Security Shift Commander is the designated person on-site qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the plant Security Plan.

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 117 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety HA2 ALERT Initiating Condition:

Other conditions exist which in the judgment of the SM/TSC Director/EOF Director warrant declaration of an Alert Operating Mode Applicability:

All Emergency Action Level:

1. Other conditions exist which in the judgment of the SM/TSC Director/EOF Director indicate that events are in process or have occurred which involve actual or likely potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Basis:

This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the SM/TSC Director/EOF Director to fall under the Alert emergency class.

PROC./WORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 118 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety HA3 ALERT Initiating Condition:

Control Room evacuation has been initiated Operating Mode Applicability:

All Emergency Action Level:

Control Roan evacuation in progress Basis:

1. With the Control Room evacuated, additional support, monitoring and direction through the Technical Support Center and/or other emergency response facilities is necessary.

Inability to establish plant control from outside the Control Room within 15 minutes will escalate this event to a Site Area Emergency.

PROCJWORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 119 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety HA4 ALERT Initiating Condition:

FIRE or EXPLOSION affecting the operability of plant safety systems required to establish or maintain safe shutdown Operating Mode Applicability:

All Emergency Action Level:

1. FIRE or EXPLOSION in any Table H1 area on either unit.

Table H1 Intake Structure Containment Auxiliary Building Aux Extension Building QCST/RWT/BWST Diesel Fuel Oil Vault Transformer Yard Turbine Building AND.

Affected system parameter indications show degraded performance or plant personnel report VISIBLE DAMAGE to permanent structures or equipment within the specified area.

Basis:

This EAL addresses a FIRE/EXPLOSION and not the degradation in performance of affected systems. System degradation is addressed in the System Malfunction EALs.

The reference to damage of systems is used to identify the magnitude of the FIRE/EXPLOSION and to discriminate against minor FIREs/EXPLOSIONs. The reference to safety systems is included to discriminate against FIREs/EXPLOSIONs in areas having a low probability of affecting safe operation. The significance here is not that a safety system was degraded but the fact .that the FIRE/EXPLOSION was large enough to cause damage to these systems.

This situation is not the same as removing equipment for maintenance that is covered by the plant's Technical Specifications. Removal of equipment for maintenance is a planned activity controlled in accordance with procedures and, as such, does not constitute a substantial degradation in the level of safety of the plant. A FIRE/EXPLOSION is an UNPLANNED activity and, as such, does constitute a substantial degradation in the level of safety of the plant. In this situation, an Alert classification is warranted.

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INORK PLAN TITLE: PAGE: 120 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety (HA4)

The inclusion of a "report of VISIBLE DAMAGE" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The occurrence of the EXPLOSION with reports of evidence of damage is sufficient for declaration. The declaration of an Alert and the activation of the Technical*

Support Center will provide the SM/TSC Director/EOF Director with the resources needed to perform these damage assessments. The SM/TSC Director/EOF Director also needs to consider any security aspects of the EXPLOSIONs, if applicable.

Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Levels/Radiological Effluent, or SM/TSC Director/EOF Director Judgment EALs.

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IWORK PLAN TITLE: PAGE: 121 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety HA5 ALERT Initiating Condition:

Release of toxic or flammable gases within or adjacent to a VITAL AREA which jeopardizes operation of systems required to establish or maintain safe shutdown Operating Mode Applicability:

All Emergency Action Level: 1 OR 2

1. Report or detection of toxic gases within or adjacent to a VITAL AREA in concentrations that may result in an atmosphere IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (IDLH).

2. Report or detection of gases in concentration greater than the LOWER FLAMMABILITY LIMIT within or adjacent to a VITAL AREA.

Basis:

This IC is based on gases that affect the safe operation of the plant. This IC applies to buildings and areas adjacent to plant VITAL AREAs or other significant buildings or areas (i.e., service water intake). The intent of this IC is not to include buildings (e.g., warehouses) or other areas that are not immediately adjacent to plant VITAL AREAs. It is appropriate that increased monitoring be done to ascertain whether consequential damage has occurred.

EAL #1 is met if measurement of toxic gas concentration results in an atmosphere that is IDLH within a VITAL AREA or any area or building adjacent to a VITAL AREA. Exposure to an IDLH atmosphere will result in immediate harm to unprotected personnel, and would preclude access to any such affected areas.

Areas that require only temporary access that can be supported by the use of respiratory protection should not be considered as exceeding this threshold.

EAL #2 is met when the flammable gas concentration in a VITAL AREA or any building or area adjacent to a VITAL AREA exceeds the LOWER FLAMMABILITY LIMIT.

Flammable gasses, such as hydrogen and acetylene, are routinely used to maintain plant systems (hydrogen) or to repair equipment/components (acetylene

- used in welding). This EAL addresses concentrations at which gases can ignite/support combustion. An uncontrolled release of flammable gasses within a facility structure has the potential to affect safe operation of the plant by limiting either operator or equipment operations due to the potential for ignition and resulting equipment damage/personnel injury. Once it has been determined that an uncontrolled release is occurring, then sampling must be done to determine if the concentration of the released gas is within this range.

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/WORK PLAN TITLE: PAGE: 122 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION l CHANGE: XXCX-XX-0l Attachment 3 Hazards and Other Conditions Affecting Plant Safety (RA5)

Escalation to a higher emergency class, if appropriate, will be based on System Malfunction, Fission Product Barrier Degradation, Abnormal Rad Levels/

Radioactive Effluent, or SM/TSC Director/EOF Director Judgment EALs.

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IWORK PLAN TITLE: PAGE: 123 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety HA6 ALERT Initiating Condition:

Natural and destructive phenomena affecting the plant VITAL AREA Operating Mode Applicability:

All Emergency Action Level: 1 OR 2 OR 3 OR 4 OR 5 OR 6

1. An earthquake is felt and the O.lg acceleration alarm annunciates indicating an Operating Basis Earthquake has occurred.

2. Tornado or high winds greater than 67 mph within PROTECTED AREA boundary resulting in VISIBLE DAMAGE to any of the plant structures/equipment in Table H2 or Control Room indication of degraded performance of those systems on either unit.

3. Vehicle crash within PROTECTED AREA boundary resulting in VISIBLE DAMAGE to any of the plant structures/equipment in Table H2 or Control Room indication of degraded performance of those systems.

4. Turbine failure-generated missiles resulting in VISIBLE DAMAGE to or penetration of any of the plant structures/equipment in Table H2 or Control Room indication of degraded performance of those systems.

5. Uncontrolled flooding in areas of the plant that results in degraded safety system performance as indicated in the control room or that creates industrial safety hazards (e.g., electric shock) that precludes access necessary to operate or monitor safety equipment.

6. Lake Dardanelle level less than 335 feet and Emergency Cooling Pond inoperable Table H2 Intake Structure Fuel Handling Building Containment Auxiliary Building QCST/RWT/BWST Diesel Fuel Oil Vault Start Up Transformer Emergency Cooling Pond Control Room

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ANORK PLAN TITLE: PAGE: 124 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety (HA6)

Basis:

These EALs escalate from the NUE EALs in HU6 in that the occurrence of the event has resulted in VISIBLE DAMAGE to plant structures or areas containing equipment necessary for a safe shutdown, or has caused damage to the safety systems in those structures evidenced by control indications of degraded system response or performance. The occurrence of VISIBLE DAMAGE and/or degraded system response is intended to discriminate against lesser events.

The initial "report" should not be interpreted as mandating a lengthy damage assessment prior to classification. No attempt is made in this EAL to assess the actual magnitude of the damage. The significance here is not that a particular system or structure was damaged, but rather, that the event was of sufficient magnitude to cause this degradation. Escalation to higher classifications occurs on the basis of other EALs (e.g., System Malfunction).

EAL #1 is based on seismic events of a magnitude that can result in a plant VITAL AREA being subjected to forces beyond design limits, and thus damage may be assumed to have occurred to plant safety. systems. See EPRI-sponsored "Guidelines for Nuclear Plant Response to an Earthquake", dated October 1989, for information on seismic event categories.

EAL #2 is based on the assumption that a tornado striking (touching down) or high winds within the PROTECTED AREA may have potentially damaged plant structures containing functions or systems required for safe shutdown of the plant. The high wind value in EAL #2 is conservatively based on the SAR design basis for Unit 1 of 67 mph. Unit 2 Design basis is 80 mph. If damage is confirmed visually or by other plant indications, escalation to Alert is appropriate.

EAL #3 is intended to address crashes of vehicle types large enough to cause significant damage to plant structures containing functions and systems required for safe shutdown of the plant. If the crash is confirmed to affect a plant VITAL-AREA, escalation to Alert is appropriate.

EAL #4 is intended to address the threat to safety related equipment imposed by missiles generated by main turbine rotating component failures. The list of areas includes all areas containing safety-related equipment, their controls, and their power supplies. This EAL is, therefore, consistent with the definition of an ALERT in that if missiles have damaged or penetrated areas containing safety-related equipment the potential exists for substantial degradation of the level of safety of the plant.

EAL #5 addresses the effect of internal flooding that has resulted in degraded performance of systems affected by the flooding, or has created industrial safety hazards (e.g., electrical shock) that preclude necessary access to operate or monitor safety equipment. The inability to operate or monitor safety equipment represents a potential for substantial degradation of the level of safety-of the plant. This flooding may have been caused by internal events such as component failures, equipment misalignment, or outage activity mishaps. The areas include those areas that contain systems required for safe shutdown of the plant that are not designed to be wetted or submerged.

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IWORK PLAN TITLE: PAGE: 125 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety (HA6)

EAL #6 addresses site specific phenomena which has the potential for the loss of primary and secondary heat sink.

Reference Documents:

1. OP-1203.025 "Natural Emergencies"

2. OP-2203.008 "Natural Emergencies"

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 126 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety.

HS1 SITE AREA EMERGENCY Initiating Condition:

Confirmed security event in a plant VITAL AREA Operating Mode Applicability:

All Emergency Action Level: 1 OR 2

1. INTRUSION into the plant VITAL AREA by a HOSTILE FORCE.

2. Other security events as determined from Safeguards Contingency Plan and reported by the Security Shift Commander.

Basis:

This class of security events represents an escalated threat to plant safety above that contained in the Alert ICs in that a HOSTILE FORCE has progressed from the PROTECTED AREA to a VITAL AREA.

Consideration should be given to the following types of events when evaluating an event against the criteria of the site specific Security Contingency Plan:

SABOTAGE and HOSTAGE/EXTORTION. The Safeguards Contingency Plan identifies numerous events/conditions that constitute a threat/compromise the Station's security. Only those events that involve actual or likely major failures of plant functions needed for protection of the public need to be considered. The following events would not normally meet this requirement: failure by a member of the security force to carry out an assigned/required duty, internal disturbances, loss/compromise of safeguards materials or strike actions.

Loss of plant control would escalate this event to a GENERAL EMERGENCY.

The Security Shift Commander is the designated person on-site qualified and trained to confirm that a security event is occurring or has occurred. Training on security event classification confirmation is closely controlled due to the strict secrecy controls placed on the plant Security Plan.

PROC.JORK PLAN NO. PROCEDURE/WORK PLANTITLE: PAGE: 127 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-.

Attachment 3 Hazards and Other Conditions Affecting Plant Safety HS2 SITE AREA EMERGENCY Initiating Condition:

Other conditions exist which in the judgment of the SM/TSC Director/EOF Director warrant declaration. of Site Area Emergency Operating Mode Applicability:

All Emergency Action Level:

1. Other conditions exist which in the judgment of the SM/TSC Director/EOF Director indicate that events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Any releases are not expected to result in exposure levels which exceed EPA Protective Action Guideline exposure levels beyond the exclusion area.

Basis:

This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the SM/TSC Director/EOF Director to fall under the emergency class description for Site Area Emergency.

PROC.JWORK PLAN NO. PROCEDURENVORK PLAN TITLE: PAGE: 128 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 .

Attachment 3 Hazards and Other Conditions Affecting Plant Safety HS3 SITE AREA EMERGENCY Initiating Condition:

Control Room evacuation has been initiated and plant control cannot be established Operating Mode Applicability:

All Emergency Action Level: 1 AND 2

1. a. Control room evacuation has been initiated.

AND

b. Control of the plant cannot be established within 15 minutes.

Basis:

Expeditious transfer of safety systems has not occurred but fission product barrier damage may not yet be indicated. The intent of this IC is to capture those events where control of the plant cannot be reestablished in a timely manner.

The determination of whether or not control is established outside of the Control Room is based on SM/TSC Director/EOF Director judgment. The SM/TSC Director/EOF Director is expected to make a reasonable, informed judgment within 15 minutes that control of the plant has or has not been established.

The intent of the EAL is to establish control of important plant equipment and knowledge of important plant parameters in a timely manner. Primary emphasis-should be placed on those components and instruments that supply protection for and information about safety functions such as reactivity control (ability to shutdown the reactor and maintain it shutdown), RCS inventory (ability to cool the core), and secondary heat removal (ability to maintain a heat sink).

Escalation of this event, if appropriate, would be by Fission Product Barrier Degradation, Abnormal Rad Levels/Radiological Effluent, or SM/TSC Director/EOF Director Judgment EALs.

PROCJWORK PLAN NO. PROCEDUREMNORK PLAN TITLE: PAGE: 129 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 Hazards and Other Conditions Affecting Plant Safety HG1 GENERAL EMERGENCY Initiating Condition:

Security event resulting in loss of physical control of the facility Operating Mode Applicability:

All Emergency Action Level:

1. A HOSTILE FORCE has taken control of plant equipment such that plant personnel are unable to operate equipment required to maintain safety functions.

Basis:

This IC encompasses conditions under which a HOSTILE FORCE has taken physical control of VITAL AREAs (containing vital equipment or controls of vital equipment) required to maintain safety functions and control of that equipment cannot be transferred to and operated from another location. These safety functions are reactivity control (ability to shut down the reactor and keep it shutdown) RCS inventory (ability to cool the core), and secondary heat removal (ability to maintain a heat sink). If control of the plant equipment necessary to maintain safety functions can be transferred to another location, then the above initiating condition is not met.

This EAL should also address loss of physical control of spent fuel pool cooling systems if imminent fuel damage is likely (e.g., freshly off-loaded reactor core in pool).

Loss of physical control of the Control Room or remote shutdown/alternate shutdown capability alone may not.prevent the ability to maintain safety functions. Design of the remote shutdown/alternate capability and the location of the transfer switches should be taken into account.

PROCJWORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 130 of 156 1903.010 EMERGENCY ACTIONI LEVEL CLASSIFICATION CHANGE: XXX-XX-0 I

- Attachment 3 Hazards and Other Conditions Affecting Plant Safety HG2 GENERAL EMERGENCY Initiating Condition:

Other conditions exist which in the judgment of the SM/TSC Director/EOF Director warrant declaration of General Emergency Operating Mode Applicability:

All Emergency Action Level:

1. Other conditions exist which in the judgment of the SM/TSC Director/EOF Director indicate that events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels beyond the exclusion area.

Basis:

This EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conditions exist which are believed by the SM/TSC Director/EOF Director to fall under the General Emergency class.

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IWORK PLAN TITLE: PAGE: 131 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX.O Attachment 3 SYSTEM MALFUNCTION Sul NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Loss of all offsite power to vital 4.16 KV busses for greater.than 15 minutes Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level:

1. Loss of power to all Unit Auxiliary and Startup Transformers on either unit for greater than 15 minutes..

Unit 1 Unit 2 SUl SU3 SU2 SU2 Unit Aux Unit Aux

AND Both vital 4.16 KV busses supplied power from independent diesel generator.

Unit 1 Unit 2 1DG1 2DG1 1DG2 2DG2 AACG AACG Basis:

Prolonged loss of AC power reduces required redundancy and potentially degrades the level of safety of the plant by rendering the plant more vulnerable to a complete loss of AC power (e.g.,.Station Blackout). Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. This EAL.is based on a failure of offsite power sources resulting in a loss of RCPs, loss of turbine load, and a loss of main feedwater. This leaves the electrical distribution system with only one or both of the vital ES busses energized.

Loss of the 6.9 KV busses and non-vital 4.16 KV busses puts the plant in a natural circulation mode with decay heat being removed by the EFW system.

Maintaining the required components for natural circulation cooling is of vital importance. Loss of any component function necessary to maintain natural circulation may require escalation.

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INORK PLAN TITLE: PAGE: 132 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0.

Attachment 3 SYSTEM -MALFUNCTION (SUI)

The EAL allows credit for operation of installed design feature (Alternate AC Diesel.Generator).

Reference Documents:

1. 1202.007, "Degraded Power"

2. 1202.008, "Blackout"

3. 2202.007, "Loss of Off-Site Power"

4. 2202.008, "Station Blackout"

5. 2104.037, "Alternate AC Diesel Generator Operations"

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JWORK PLAN TITLE: PAGE: 133 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALFUNCTION SU2 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Inability to reach required shutdown within Technical Specification limits Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level:

1. Plant is not brought to required operating mode within Technical Specifications LCO action statement time.

Basis:

Limiting Conditions for Operation (LCOs) require the plant to be brought to a required shutdown mode when the Technical Specification required configuration cannot be restored. Depending on the circumstances, this may or may not be an emergency or precursor to a more severe condition. In any case, the initiation of plant shutdown required by the site Technical Specifications requires a four hour report under 10 CFR 50.72 (b) Non-emergency events. The plant is within its safety envelope when being shut down within the allowable action statement time in the Technical Specifications. An immediate NUE is required when the plant is not brought to the required operating mode within the allowable action statement time in the Technical Specifications. Declaration of a NUE is based on the time at which the LCO-specified action statement time period elapses under the site Technical Specifications and is not related to how long a condition may have existed. Other required Technical Specification shutdowns that involve precursors to more serious events are addressed by other System Malfunction, Hazards, or Fission Product Barrier Degradation ICs.

Reference Documents:

1. AN02 Technical Specifications

2. ANOI Technical Specifications

PROC.MORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 134 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALFUNCTION SU3 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

UNPLANNED loss of most or all safety system annunciation or indication in the Control Room for greater than 15 minutes Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level: 1 OR 2

1. UNPLANNED loss of annunciators or indicators associated with safety systems for greater than 15 minutes as follows:

Unit 1: Loss of AC AND DC to greater than or equal to 50% of Control Room annunciators Unit 2: Loss of AC AND DC to greater than or equal to 9 Control Room annunciator panels OR

2. UNPLANNED loss of 75% of indicators associated with safety systems for greater than 15 minutes.

Basis:

This IC and its associated EALs are intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment.

Recognition of the availability of computer based indication equipment is considered (e.g., SPDS, plant computer, etc.).

Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is a higher risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions.

It is further recognized that each plant design provides redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a

PROC.MVORK PLAN NO. PROCEDURENVORK PLAN TITLE: PAGE: 135 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALFUNCTION (SU3) large portion of indications, the concern is included in this EAL .due to difficulty associated with assessment of plant conditions.. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50.72. If the shutdown is not in compliance with the Technical Specification action, the NUE is based on SU2.

Annunciators or indicators for this EAL must include those identified in the Abnormal Operating Procedures, in the Emergency Operating Procedures, and in other EALs (e.g., area, process, and/or effluent rad monitors, etc.). The loss of control room annunciators increases the difficulty to recognize changing plant conditions. It is estimated that if approximately 75% of the safety system annunciators or indications are lost, there is an increased risk that a degraded plant condition could go undetected. For ANO2 the selection of 9 annunciator panels was chosen since if greater than 9 annunciator panels were lost this would mean that all AC and DC was lost to either the Red or Green safety system. Any less than 9 annunciator panels would mean that a localized problem exists that does not affect the annunciators for an entire train.

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

Due to the limited number of safety systems in operation during cold shutdown, refueling, and.defueled modes, no EAL is indicated during these modes of operation.

This NUE will be escalated to an Alert if a transient is in progress during the loss of annunciation or indication (SA4).

Basis Documents:

1. 1203.043, "Loss Control Room Annunciator"

2. 2203.042, "Loss of Annunciators"

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Attachment 3 SYSTEM MALFUNCTION SU4 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Fuel clad degradation Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level:

1. RCS sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits.

Unit 1:

RCS Sample Analysis: greater than 3.50 pCi/gm IDE RCS Sample Analysis: greater than 72/E pCi/gm Gross Activity Unit 2:

RCS Sample Analysis: greater than 1.0 pCi/gm IDE RCS Sample Analysis: greater than 100/Et pCi/gm Gross Activity Basis:

This IC and its associated EALs are included as an NUE because it is considered to be a potential degradation in the level of safety of the plant and a potential precursor of more serious problems. EAL #1 addresses coolant samples exceeding coolant technical specifications for iodine spike. Escalation of this EAL to the Alert level is via the Fission Product Barrier Degradation Monitoring EALs. Though the referenced Technical Specification limits are mode dependent, it is appropriate that the EALs be applicable in all modes, as they indicate a potential degradation in the level of safety of the plant. The companion EAL to SU4 for the Cold Shutdown/Refueling modes is CU5.

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/WORK PLAN TITLE: PAGE: 137 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALFUNCTION SU5 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

RCS leakage Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level: 1 OR 2

1. Unidentified or pressure boundary leakage greater than 10 gpm.

2. Identified leakage greater than 25 gpm.

Basis:

This IC is included as an NUE because the condition may be a precursor of more serious conditions and, as result, is considered to be a potential degradation of the level of safety of the plant. The 10 gpm value for the unidentified and pressure boundary leakage was selected as it is observable with normal Control Room indications. Lesser values must generally be determined through time-consuming surveillance tests (e.g., mass balances). The EAL for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage. In either case, escalation of this IC to the Alert level is via FA1.

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/WORK PLAN TITLE: PAGE: 138 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 .

Attachment 3 SYSTEM MALFUNCTION SU6 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

UNPLANNED loss of all onsite or offsite communications capabilities Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level: 1 OR 2

1. Loss of all onsite communications capability (Table Ml) affecting the ability to perform routine operations.

Table Ml Onsite Communications Equipment Station radio system Plant paging system In-plant telephones Plant cell phones Gaitronics

2. Loss of all offsite communications capability (Table M2)

Table M2 Offsite Communications Equipment All telephone lines (commercial

-and microwave)

Station radio system ENS Cellular phones Basis:

The purpose of this IC and its associated EALs is to recognize a loss of communications capability that either defeats the plant operations staff's ability to perform routine tasks necessary for plant operations or the ability to communicate problems with offsite authorities. The loss of offsite communications ability is expected to be significantly more comprehensive than the condition addressed by 10 CFR 50.72.

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/WORK PLAN TITLE: PAGE: 139 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALFUNCTION (SU6)

The availability of one method of ordinary offsite communications .is sufficient to inform state and local authorities of plant problems. This EAL is-intended to be used only when extraordinary means (e.g., relaying of information from radio transmissions, individuals being sent to offsite locations, etc.) are being utilized to make communications possible.

Basis Documents:

1. 1903.062, "Communications System Operating Procedure"

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IWORK PLAN TITLE: PAGE: 140 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION

. . ~CHANGE: XXX-XX-0O Attachment 3 SYSTEM MALFUNCTION SU8 NOTIFICATION OF UNUSUAL EVENT Initiating Condition:

Inadvertent criticality OPERATING MODE APPLICABILITY:

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level:

1. An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.

Unit 1:

Greater than 2 DPM (Source Range)

Greater than 3 DPM (Intermediate Range)

Unit 2:

Greater than 1.6 DPM Basis:

This IC addresses inadvertent criticality events. While the primary concern is criticality events that occur in cold shutdown or refueling modes (NUREG 1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States), this IC is applicable in other modes in which inadvertent criticalities are possible. This IC indicates a potential degradation of the level of safety of the plant, warranting a NUE classification. This IC excludes inadvertent criticalities that occur during planned reactivity changes associated with reactor startups (e.g., criticality earlier than estimated).

The Cold Shutdown/Refueling IC is CU8.

This condition can be identified using the startup rate monitor. The term "sustained" is used in order to allow exclusion of expected short term positive startup rates from planned control rod movements such as shutdown bank withdrawal. These short term positive startup rates are the result of the rise in neutron population due to subcritical multiplication.

Escalation would be by the fission product barrier EALs, as appropriate to the operating mode at the time of the event, or by SM/TSC Director/EOF Director Judgment.

Reference Documents:

1. 1203.012G, "Annunciator K08 Corrective Action"

2. 2203.012D, "Annunciator 2K04 Corrective Action"

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IWORK PLAN TITLE: PAGE: 141 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALFUNCTION SA2 ALERT Initiating Condition:

Failure of Reactor Protection System instrumentation to complete or initiate an automatic reactor trip once a Reactor Protection System setpoint has been exceeded and manual trip was successful Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Emergency Action Level:

1. Indication(s) exist that indicate that reactor protection system setpoint was exceeded and automatic trip-did not occur, and a successful manual trip or DSS trip occurred.

Basis:

This condition indicates failure of the reactor protection system to trip the reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not function in response to a plant transient and thus the plant safety has been compromised, and design limits of the fuel may have been exceeded. An Alert is indicated because conditions exist that lead to potential loss of fuel clad or RCS barriers. Reactor protection system setpoint being exceeded, rather than limiting safety system setpoint being exceeded, is specified here because failure of the reactor protection system is the issue. A manual trip is any set of actions by the reactor operator(s) at the reactor control console which causes control rods to be rapidly inserted into the core and brings the reactor subcritical (e.g., manual reactor trip, diverse trip initiation, de-energizing rod drive mechanisms). Failure of manual trip would escalate the event to a Site Area Emergency (SS2).

The operator may not detect the RPS failure prior to performing the manual trip. The failure would be detected by reviewing the post trip sequence of events printout from the plant computer and the emergency class would be declared, at that time.

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IWORK PLAN TITLE: PAGE: 142 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 .

CHNE.XXX-

- Attachment 3 SYSTEM MALFUNCTION SA4 ALERT Initiating Condition:

UNPLANNED loss of most or all safety system annunciation or indication in Control Room with either (1) a PLANT TRANSIENT in progress, or (2) SPDS and PMS dynamic alarm functions are unavailable Operating. Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level:

1. UNPLANNED loss of annunciators or indicators associated with safety systems for greater than 15 minutes as follows:

Unit 1:

Loss of AC AND DC to greater than or equal to 50% of Control Room annunciators Unit 2:

Loss of AC AND DC to greater than or equal to 9 Control Room annunciator panels AMD Either of the following: (a or b)

a. PLANT TRANSIENT is in progress.

OR

b. SPDS and PMS dynamic alarm functions are unavailable.

Basis:

This EAL is intended to recognize the difficulty associated with monitoring changing plant conditions without the use of a major portion of the annunciation or indication equipment during a transient. Recognition of the availability of computer based indication equipment is considered (e.g., SPDS, plant computer, etc.).

"Planned" loss of annunciators or indicators includes scheduled maintenance and testing activities.

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IWORK PLAN TITLE: PAGE: 143 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALFUNCTION (SA4)

Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is higher risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Manager be tasked with making a judgment decision as to whether additional personnel are required to provide more monitoring of system operation.

It is further recognized that most plant designs provide redundant safety system indication powered from separate uninterruptible power supplies. While failure of a large portion of annunciators is more likely than a failure of a large portion of indications, the concern is included in this EAL due to difficulty associated with assessment of plant conditions. The loss of specific, or several, safety system indicators should remain a function of that specific system or component operability status. This will be addressed by the specific Technical Specification. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50.72. If the shutdown is not in compliance with the Technical Specification action, the NUE is based on SU2 Due to the limited number of safety systems in operation during cold shutdown, refueling and defueled modes, no ZAL is indicated during these modes of operation.

This Alert will be escalated to a Site Area Emergency (SS6) if the operating crew cannot monitor the transient in progress.

Reference Documents:

1. 1015.037, "Post Transient Review"

2. 1203.043, "Loss Control Room Annunciator"

3. 2203.042, "Loss of Annunciators"

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IWORK PLAN TITLE: PAGE: 144 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XX-XX-0 Attachment 3 SYSTEM MALFUNCTION SA5 ALERT Initiating Condition:

AC power capability to vital 4.16 KV busses reduced to a single power source for greater than 15 minutes such'that any additional single failure would result in station blackout Operating. Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level:

1. Only ONE vital 4.16 KV bus energized from a single power source for greater than 15 minutes.

Unit 1 Unit 2 A3 2A3 A4 2A4 AND Any additional single failure will result in station blackout.

Basis:

This IC and its associated EAL is intended to provide an escalation from SU1, "Loss of All Offsite Power To Vital 4.16 KV Busses for Greater Than 15 Minutes." The condition indicated is the degradation of the offsite and onsite power systems such that any additional single failure would result in a station blackout. This condition could occur due to a loss of offsite power with a concurrent failure of one emergency generator to supply power to its emergency busses. Another related condition could be the loss of all offsite power and loss of onsite emergency diesels with only one train of emergency busses being backfed from the unit main generator, or the loss of onsite emergency diesels with only one train of emergency busses being backfed from offsite power. The subsequent loss of this single power source would escalate the event to a Site Area Emergency in accordance with SS1 Loss of the 6.9 KV busses and non-vital 4.16 KV busses puts the plant in a natural circulation mode with decay heat being removed by the EFW System.

Maintaining the required components for natural circulation cooling is of vital importance.

The EAL allows credit for operation of installed design feature (Alternate AC Diesel Generator).

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/WORK PLAN TITLE: PAGE: 145 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-O Attachment 3 SYSTEM MALFUNCTION (SA5)

Even though a vital 4.16 KV bus may be energized, if necessary loads (i.e.,

loads that if lost would inhibit decay heat removal capability or reactor vessel makeup capability) are not operable on the energized bus then the bus should not be considered operable. If this bus was the only energized bus then a Site Area Emergency per SS1 should be declared.

Reference Documents:

1. 1202.007, "Degraded Power"

2. 1202.008, "Blackout"

3. 2202.007, "Loss of Off-Site Power"

4. 2202.008, "Station Blackout"

5. 2104.037, "Alternate AC Diesel Generator Operations"

PROCJWORK PLAN NO. PROCEDUREJWORK PLAN TITLE: PAGE: 146 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 K

Attachment 3 SYSTEM MALFUNCTION SS1 SITE AREA EMERGENCY Initiating Condition:

Loss of all offsite power and loss of all onsite AC power to vital 4.16 KV busses Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level:

1. Loss of power to ALL Unit Auxiliary and Startup transformers on a unit.

Unit . Unit 2 SUl SU3 SU2 SU2 Unit Aux Unit Aux AND NO vital 4.16 KV bus .being supplied power from ANY diesel generator for greater than 15 minutes.

Unit 1 DG Unit 2 DG Unit 1 Bus Unit 2 Bus 1DGI 2DG1 A3 2A3 1DG2 2DG2 A4 2A4 AACG AACG Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including DHR or SDC, ECCS, containment heat removal and the ultimate heat sink. Prolonged loss of all AC power will cause core uncovering and loss of containment integrity, thus this event can escalate to a General Emergency.

The 15 minute duration is selected to exclude transient or momentary power losses.

Escalation to General Emergency is via fission product barrier degradation FG1 or SG1.

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/WORK PLAN TITLE: PAGE: 147 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-O Attachment 3 SYSTEM MALFUNCTION (SS1)

Loss of the 6.9 KV busses and non-vital 4.16 KV busses puts the plant in a natural circulation mode with Decay Heat being removed by the EFW System.

Maintaining the required components for Natural Circulation Cooling is of vital importance. Consideration should be given to operable loads necessary to remove decay heat or provide Reactor Vessel makeup capability when evaluating loss of AC power to vital 4.16 KV busses. Even though a vital bus may be energized, if necessary loads (i.e., loads that if lost would inhibit decay heat removal capability or Reactor Vessel makeup capability) are not operable on the energized bus, then the bus should not be considered operable for this IC. If this bus was the only energized bus, than a Site Area Emergency per SS1 should be declared.

Reference Documents:

1. 1202.007, Degraded Power

2. 1202.008, Blackout

3. 2202.007, Loss of Off-Site Power

4. 2202.008, Station Blackout

5. 2104.037, Alternate AC Diesel Generator Operations

PROCJWORK PLAN NO. PROCEDUREJWORK PLAN TITLE: PAGE: 148 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALFUNCTION SS2 SITE AREA EMERGENCY Initiating Condition:

Failure of Reactor Protection System instrumentation to complete or initiate an automatic reactor trip once a Reactor Protection System setpoint has been exceeded and manual trip .was NOT successful Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Emergency Action Level:

1. Indication(s) exist that automatic and manual reactor trips were not successful.

Basis:

Automatic and manual trip are not considered successful if action away from the reactor control console was required to trip the reactor.

Under these conditions, the reactor is producing more heat than the maximum decay heat load for which the safety systems are designed. A Site Area Emergency is indicated because conditions exist that lead to imminent loss or potential loss of both fuel clad and RCS barriers. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response. Escalation of this event to a General Emergency would be via FG1 or HG2.

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 149 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALFUNCTION SS3 SITE AREA EMERGENCY Initiating Condition:

Loss of all vital DC power Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level:

1. Loss of ALL of the following busses has occurred for greater than 15 minutes:

Unit 1: Unit 2:

DOI and D02 2D01 and 2D02 Basis:

Battery bus voltage indicating less than 105 volts constitutes loss-of DC associated busses. Loss of all DC power compromises ability to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is significant decay heat and sensible heat in the reactor system. Escalation to a General Emergency would occur via AGi or FG1. Fifteen minutes was selected as a threshold to exclude transient or momentary power losses.

PROCJNORK PLAN NO. PROCEDUREIWORK PLAN TITLE: PAGE: 150 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 .

Attachment 3 SYSTEM MALFUNCTION SS4 SITE AREA EMERGENCY Initiating Condition:

Complete loss of heat removal capability Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level:

1. Loss of core cooling and heat sink as indicated by:

a. Loss of ALL Normal Feedwater AND

b. Loss of ALL Emergency/Auxiliary Feedwater

c. High Pressure Injection (Unit 1)/Once-Through Core Cooling (Unit 2)

NOT established.

Basis:

This EAL addresses complete loss of functions, including ultimate heat sink, required for hot shutdown with the reactor at pressure and temperature.

Reactivity control is addressed in other EALs.

Under these conditions, there is an actual major failure of a system intended for protection of the public. Thus, declaration of a Site Area Emergency is warranted. Escalation to General Emergency would be via AG1 or FG1.

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INORK PLAN TITLE: PAGE: 151 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-O Attachment 3 SYSTEM MALFUNCTION SS6 SITE AREA EMERGENCY Initiating Condition:

Inability to monitor a TRANSIENT in progress Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level:

l.a. Loss of most or all annunciators associated with safety systems.

Unit 1: Loss of greater than or equal to 50% of Control Room Annunciators Unit 2: Loss of AC AND DC to greater than or equal to 9 Control Room Annunciator panels AND

b. SPDS and PMS dynamic alarm functions are unavailable.

AND

c. Loss of 75% of indicators associated with safety systems.

AND

d. A TRANSIENT in progress.

Basis:

This IC and it associated EAL is intended to recognize the inability of the Control Room staff to monitor the plant response to a transient. A Site Area Emergency is considered to exist if the control room staff cannot monitor safety functions needed for protection of the public.

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IWORK PLAN TITLE: PAGE: 152 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION

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Attachment 3 SYSTEM MALFUNCTION (SS6)

Indications needed to monitor safety functions necessary for protection of the public must include Control Room indications, computer generated indications and dedicated annunciation capability. The specific indications should be those used to determine such functions as the ability to shut down the reactor, maintain the core cooled, to maintain the reactor coolant system intact, and to maintain containment intact (FS1, FG1).

"Planned" and "UNPLANNED" actions are not differentiated since the loss of instrumentation of this magnitude is of such significance during a transient that the cause of the loss is not an ameliorating factor.

Quantification of "Most" is arbitrary, however, it is estimated that if approximately 75% of the safety system annunciators or indicators are lost, there is a higher risk that a'degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgment threshold for determining the severity of the plant conditions. It is also not intended that the Shift Manager be tasked with making a judgment decision as to whether additional personnel are required to provide more monitoring of system operation.

Reference Documents:

1. 1015.037, "Post Transient Review"

2. 1203.043, "Loss Control Room Annunciator"

3. 2203.042, "Loss of Annunciators"

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MWORK PLAN TITLE: PAGE: 153 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALFUNCTION SG1 GENERAL EMERGENCY Initiating Condition:

Prolonged loss of all offsite power and prolonged loss of all onsite AC power to vital 4.16 KV busses Operating Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Hot Standby (Mode 3)

Hot Shutdown (Mode 4)

Emergency Action Level:

1. Loss of power to all unit auxiliary and startup transformers on a unit.

Unit 1 Unit 2 SUl SU3 SU2 SU2 Unit Aux Unit Aux AND NO vital 4.16 KV bus being supplied power from ANY diesel generator.

Unit 1 DG Unit 2 DG Unit 1 Bus Unit 2 Bus lDG1 2DG1 A3 2A3 1DG2 2DG2 A4 2A4 AACG AACG AND Either of the following: (a or b)

a. Restoration of at least one emergency bus within four (4) hours is not likely OR

b. FA1 entry conditions met.

PROCJWORK PLAN NO. PROCEDURE/WORK PLAN TITLE: PAGE: 154 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALEUNCTION (SG1)

Basis:

Loss of all AC power compromises all plant safety systems requiring electric power including DHR, SDC, ECCS, containment heat removal and the ultimate heat sink. Prolonged loss of all AC power will lead to loss of fuel clad, RCS, and containment barriers. The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore AC power is based on the results of the calculations referenced below. Appropriate allowance for offsite emergency response including evacuation of surrounding areas should be considered. Although this IC may be viewed as redundant to the Fission Product Barrier Degradation ICs, its inclusion is necessary to better assure timely recognition and emergency response.

This IC is specified to assure that in the unlikely event of a prolonged station blackout, timely recognition of the seriousness of the event occurs and that declaration of a General Emergency occurs as early as is appropriate, based on a reasonable assessment of the event trajectory.

The likelihood of restoring at least one emergency bus should be based on a realistic appraisal of the situation since a delay in an upgrade decision based on only a chance of mitigating the event could result in a loss of valuable time in preparing and implementing public protective actions.

In addition, under these conditions, fission product barrier monitoring capability may be degraded. Although it may be difficult to predict when power can be restored, it is necessary to give the SM/TSC Director/EOF Director a reasonable idea of how quickly the need to declare a General Emergency may be based on two major considerations:

1. Are there any present indications that core cooling is already degraded to the point that Loss or Potential Loss of fission product barriers is imminent.

2. If there are no present indications of such core cooling degradation, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential loss of the third barrier can be prevented.

Thus, indication of continuing core cooling degradation must be based on fission product barrier monitoring with particular emphasis on SM/TSC Director/EOF Director judgment as it relates to imminent Loss or Potential Loss of fission product barriers and degraded ability to monitor fission product barriers.

Reference Documents:

1. ANO-1 Calculation 85-E-0072-02, "Time from Loss of All AC Power to Loss of Subcooling"

2. ANO-2 Calculation 85-E-0072-01, "Time from Loss of All AC Power to Loss of Subcooling"

PROC.IWORK PLAN NO. PROCEDUREMWORK PLAN TITLE: PAGE: 155 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALFUNCTION SG2 GENERAL EMERGENCY Initiating Condition:

Failure of the Reactor Protection System to complete an automatic trip and manual trip was NOT successful and there is indication of an extreme challenge to the ability to cool the core Operating .Mode Applicability:

Power Operation (Mode 1)

Startup (Mode 2)

Emergency Action Level:

1. Indications exist that automatic and manual reactor trips were NOT successful.

AND Either of the following: (a or b)

a. Indication(s) exists that core cooling is extremely challenged.

Unit 1: OUTSIDE Region 1 of EOP Figure 4 Unit 2: CET average temperature greater than 700'F OR

b. Indication(s) exist that heat removal is extremely challenged with ALL of the following being TRUE:

Loss of ALL normal Feedwater

Loss of ALL Emergency/Auxiliary Feedwater

Unit 1: High Pressure Injection NOT established Unit 2: Once-Through Core Cooling NOT established Basis:

Automatic and manual trip are not considered successful if action away from the reactor control console is required to trip the reactor.

Under the conditions of this IC and its associated EALs, the efforts to bring the reactor subcritical have been unsuccessful and, as a result, the reactor is producing more heat than the maximum decay heat load for which the safety systems were designed. Although there are capabilities away from the reactor control console, such as emergency boration, the continuing temperature rise indicates that these capabilities are not effective. This situation could be a precursor for a core melt sequence.

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IWORK PLAN TITLE: PAGE: 156 of 156 1903.010 EMERGENCY ACTION LEVEL CLASSIFICATION CHANGE: XXX-XX-0 Attachment 3 SYSTEM MALFUNCTION (SG2)

The extreme challenge to the ability to cool the core is intended to mean that the core exit temperatures are at or approaching 1200*F or that the reactor vessel water level is below the top of active fuel.

Another consideration is the inability to initially remove heat during the early stages of this sequence. If emergency feedwater flow is insufficient to remove the amount of heat required by design from at least one steam generator, an extreme challenge should be considered to exist.

In the event either of these challenges exist at a time that the reactor has not been brought below the power associated with the safety system design (typically 3 to 5% power) a core melt sequence exists. In this situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended to be anticipatory of the fission product barrier matrix declaration (FG1) to permit maximum offsite intervention time.

Attachment 4 OCAN020407 Arkansas Nuclear One Deviations and Differences from the NEI 99-01, Revision 4 Emergency Action Levels

Entergy ARKANSAS NUCLEAR ONE DEVIATIONS AND DIFFERENCES FROM NEI 99-01, REV 4 EMERGENCY ACTION LEVELS

ANO NEI EAL Deviations and Differences Pagel1 of 86

GENERAL COMMENT

S ON DIFFERENCES AND DEVIATIONS:

ANO uses formatting such as ALL CAPS, bold and underline to aid the user in applying these EALs; particularly to set apart units, time frames or quality of a value or data (such as the term "valid"). Formatting choices may also involve minor grammatical differences between the ANO EALs and NEI 99-01 such as "that exceeds" vice 'exceeding", use of "If, then" statements for conditional statements, or the use of symbols (>,<). Such formatting differences between the ANO EALs and NEI 99-01 will not be noted in this document as differences or deviations when they represent format choices alone and do not change the intent or materially change the content of NEI 99-01 Initiating Conditions or EALs.

At ANO, the terms "Notification of Unusual Event", "NUE', "Unusual Event" and "UE" are used interchangeably. The term "NOUE" is not used at ANO.

At ANO, all Radiological Effluent Technical Specifications are included in the ODCM, thus "ODCM" is used in place of Technical Specifications references.

"SM/TSC Director/EOF Director" is used instead of "Emergency Director'.

"Trip" is used instead of "scram".

'Safeguards Contingency Plan" is the term used to encompass all security plans/documents.

Other words were substituted for "increase" or "decrease" such as "rise", "rising",

"elevated", "lowering', "dropping", etc. These substitutions were used in ICs and EALs.

Arkansas Nuclear One used the following definitions:

Deviation: An instance in which ANO elects not to implement one or more NEI 99-01 EALs, proposes an EAL not found in NEI 99-01, or changes an NEI 99-01 IC or EAL where such an action is not stated or implied as an option in the NEI document and the action by ANO results in substantial differences in the intent of the IC or resulting classifications using the IC.

Example: Using a factor of 300, 100, 50, or 20 in AA1 vice the factor of 200 would be a deviation because it is not only different from the NEI factor, but would result in classification differences.

Example: Changing FUl to "any loss or potential loss of any barrier" vice a loss of the containment barrier only is a deviation because it changes the intent of the IC.

ANO NEI EAL Deviations and Differences Page 2 of 86 Example: ANO did not include an EAL for plant perimeter radiation monitors because ANO does not have these monitors. This is not a deviation because NEI 99-01 specifically refers to the EAL parenthetically as "for sites having telemetered perimeter monitors."

ANO identified no deviations from NEI 99-01 in this proposed EAL scheme.

Difference: Instances in which the ANO and NEI 99-01 corresponding IC or EAL are different. In some cases, ANO may have substantially changed the wording of the ICor EAL, but the intent or even the specific application of NEI 99-01 was retained, just in a different presentation style and it is not believed that different classifications would result between the two systems.

ANO NEI EAL Deviations and Differences Page 3 of 86 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AU1 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds Two Times the Radiological Effluent Technical Specifications for 60 Minutes or Longer.

Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5)

1. VALID reading on any effluent monitor that exceeds two times the alarm setpoint established by a current radioactivity discharge permit for 60 minutes or longer.

2. VALID reading on one or more of the following radiation monitors that exceeds the reading shown for 60 minutes or longer:

(site-specific list)

3. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates, with a release duration of 60 minutes or longer, in excess of two times (site-specific technical specifications).

4. VALID reading on perimeter radiation monitoring system greater than 0.10 mR/hr above normal background sustained for 60 minutes or longer [for sites having telemetered perimeter monitors].

5. VALID indication on automatic real-time dose assessment capability greater than (site-specific value) for 60 minutes or longer [for sites having such capability].

Differences:

1. EAL #5 of NEI 99-01, Rev. 4 was renumbered EAL #4 for ANO's EALs.

2. ANO's EAL #4 does not use "60 minutes or longer" as stated by NEI 99-01 Rev. 4. RDACS (a real-time dose assessment system) uses a 60 minute rolling calculation.

ANO NEI EAL Deviations and Differences Page 4 of 86

3. ANO has no perimeter radiation monitoring system, thus EAL #4 of NEI 99-01 Rev. 4 is not applicable.

Deviations:

None.

ANO NEI EAL Deviations and Differences Page5 of 86 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Unexpected Increase in Plant Radiation.

Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)

1. a. VALID (site-specific) indication of uncontrolled water level decrease in the reactor refueling cavity, spent fuel pool, or fuel transfer canal with all irradiated fuel assemblies remaining covered by water.

AND

b. Unplanned VALID (site-specific) Direct Area Radiation Monitor reading increases

2. Unplanned VALID Direct Area Radiation Monitor readings increases by a factor of 1000 over normal* levels.

Normal levels can be considered as the highest reading in the past twenty-four hours excluding the current peak value.

Differences:

1. The word udirect" was not used at ANO for EAL #1 or #2 of ANO's EALs.

ANO terminology uses "area radiation monitors" instead of "direct area radiation monitors".

2. Reworded EAL #1 for ANO terminology (e.g., refueling canal instead of reactor refueling cavity).

Deviations:

None

ANO NEI EAL Deviations and Differences Page 6 of 86 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AA1 Initiating Condition -- ALERT Any UNPLANNED Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200 Times the Radiological Effluent Technical Specifications for 15 Minutes or Longer.

Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5)

1. VALID reading on any effluent monitor that exceeds 200 times the alarm setpoint established by a current radioactivity discharge permit for 15 minutes or longer.

2. VALID reading on one or more of the following radiation monitors that exceeds the reading shown for 15 minutes or longer:

(site-specific list)

3. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates, with a release duration of 15 minutes or longer, in excess of 200 times (site-specific technical specifications).

4. VALID reading on perimeter radiation monitoring system greater than 10.0 mR/hr above normal background sustained for 15 minutes or longer [for sites having telemetered perimeter monitors].

5. VALID indication on automatic real-time dose assessment capability greater than (site-specific value) for 15 minutes or longer [for sites having such capability].

Differences:

1. EAL # 5 of NEI 99-01 Rev. 4 was renumbered EAL #4 of ANO's EALs.

EAL #4 of ANO's EALs does not use "15 minutes" since the current real-time dose assessment program uses a rolling average calculation.

2. ANO has no perimeter radiation monitoring system, thus EAL #4 of NEI 99-01 Rev. 4 is not applicable.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 7 of 86 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AA2 Initiating Condition -- ALERT Damage to Irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of Irradiated Fuel Outside the Reactor Vessel.

Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)

1. A VALID (site-specific) alarm or reading on one or more of the following radiation monitors: (site-specific monitors)

Refuel FloorArea Radiation Monitor Fuel Handling Building Ventilation Monitor Refueling Bridge Area Radiation Monitor

2. Water level less than (site-specific) feet for the reactor refueling cavity, spent fuel pool and fuel transfer canal that will result in irradiated fuel uncovering.

Differences:

ANO used "water level drop in the refueling canal or spent fuel pool exceeds makeup capacity" in lieu of a specific water level as described in NEI 99-01 Rev.

4 guidance for EAL #2. EAL #2 of NEI 99-01 Rev. 4 was also reworded to fit ANO terminology.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 8 of 86 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AA3 Initiating Condition -- ALERT Release of Radioactive Material or Increases in Radiation Levels Within the Facility That Impedes Operation of Systems Required to Maintain Safe Operations or to Establish or Maintain Cold Shutdown Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)

1. VALID (site-specific) radiation monitor readings GREATER THAN 15 mR/hr in areas requiring continuous occupancy to maintain plant safety functions:

(Site-specific) list

2. VALID (site-specific) radiation monitor readings GREATER THAN <site specific> values in areas requiring infrequent access to maintain plant safety functions.

(Site-specific) list Differences:

For EAL #1 and #2 of ANO's EALs, a site specific list is not provided since the possible plant conditions and configurations are very diverse. The SMITSC Director/EOF Director will have to take into consideration the plant configuration and the ability to access areas necessary to maintain safe operation or perform a safe shutdown.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 9 of 86 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AS1 Initiating Condition -- SITE AREA EMERGENCY Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 100 mR TEDE or 500 mR Thyroid CDE for the Actual or Projected Duration of the Release.

Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2 or 3 or 4)

Note: If dose assessment results are available at the time of declaration, the classification should be based on EAL #2 instead of EAL #1.While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated / completed in order to determine if the classification should be subsequently escalated.

1. VALID reading on one or more of the following radiation monitors that exceeds or is expected to exceed the reading shown for 15 minutes or longer:

(site-specific list)

2. Dose assessment using actual meteorology indicates doses greater than 100 mR TEDE or 500 mR thyroid CDE at or beyond the site boundary.

3. A VALID reading sustained for 15 minutes or longer on perimeter radiation monitoring system greater than 100 mR/hr. [for sites having telemetered perimeter monitors]

4. Field survey results indicate closed window dose rates exceeding 100 mR/hr expected to continue for more than one hour; or analyses of field survey samples indicate thyroid CDE of 500 mR for one hour of inhalation, at or beyond the site boundary.

Differences:

1. Child thyroid was used for EAL #2 and #3of ANO's EALs instead of CDE used in NEI 99-01 Rev. 4. Child thyroid is more conservative than CDE.

RDACS is designed for child thyroid calculation.

2. EAL #4 in NEI 99-01 Rev. 4 was renumbered EAL #3 in ANO's EALs.

ANO NEI EAL Deviations and Differences Page 10 of 86

3. EAL #3 of NEI 99-01 Rev. 4 was not used at ANO. ANO has no perimeter radiation monitoring system.

Deviations:

None.

ANO NEI EAL Deviations and Differences Page II of 86 ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT AG1 Initiating Condition -- GENERAL EMERGENCY Offsite Dose Resulting from an Actual or Imminent Release of Gaseous Radioactivity Exceeds 1000 mR TEDE or 5000 mR Thyroid CDE for the Actual or Projected Duration of the Release Using Actual Meteorology.

Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2 or 3 or 4)

Note: If dose assessment results are available at the time of declaration, the classification should be based on EAL #2 instead of EAL #1.While necessary declarations should not be delayed awaiting results, the dose assessment should be initiated / completed in order to determine if the classification should be subsequently escalated.

1. VALID reading on one or more of the following radiation monitors that exceeds or expected to exceed the reading shown for 15 minutes or longer:

(site-specific list)

2. Dose assessment using actual meteorology indicates doses greater than 1000 mR TEDE or 5000 mR thyroid CDE at or beyond the site boundary.

3. A VALID reading sustained for 15 minutes or longer on perimeter radiation monitoring system greater than 1000 mR/hr. [for sites having telemetered perimeter monitors]

4. Field survey results indicate closed window dose rates exceeding 1000 mR/hr expected to continue for more than one hour; or analyses of field survey samples indicate thyroid CDE of 5000 mR for one hour of inhalation, at or beyond site boundary.

Differences:

1. Child thyroid was used for EAL #2 and #3 of ANO's EALs instead of CDE used in NEI 99-01 Rev. 4. Child thyroid is more conservative than CDE.

RDACS is designed for child thyroid calculation.

ANO NEI EAL Deviations and Differences Page 12 of 86

2. EAL #3 of NEI 99-01 Rev. 4 was not used at ANO. ANO has no perimeter radiation monitoring system.

3. EAL #4 of NEI 99-04 was renumbered EAL #3 in ANO's EALs.

Deviations:

None.

ANO NEI EAL Deviations and Differences Page 13 of 86 COLD SHUTDOWNIREFUELING SYSTEM MALFUNCTION cul Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Reactor Coolant System Leakage Operating Mode Applicability: Cold Shutdown Emergency Action Levels: (1 or 2)

1. Unidentified or pressure boundary leakage greater than 10 gpm.

2. Identified leakage greater than 25 gpm.

Differences:

None Deviations:

None

- ANO NEI EAL Deviations and Differences Page 14 of 86 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Reactor Coolant System Leakage Operating Mode Applicability: Refueling Emergency Action Levels: (1 or 2)

1. UNPLANNED RCS level decrease below the RPV flange for > 15 minutes

2. a. Loss of RPV inventory as indicated by unexplained {site-specific} sump and tank level increase AND

b. RPV level cannot be monitored Differences:

ANO re-worded the IC to be consistent with the wording in CA2.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 15 of 86 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU3 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Loss of all offsite power to Essential busses for greater than 15 minutes.

Operating Mode Applicability: Cold Shutdown Refueling Emergency Action Levels:

1. a. Loss of power to (site-specific) transformers for greater than 15 minutes.

AND

b. At least (site-specific) emergency generators are supplying power to emergency busses.

Differences:

1. EAL #1b of ANO's EALs was reworded for ANO terminology.

2. Initiating Condition of CU3 in ANO's EALs was reworded to use "vital" instead of uessential".

3. ANO chose to apply this IC in a "defueled" condition as well as in cold shutdown and refueling.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 16 of 86 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU4 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT UNPLANNED loss of decay heat removal capability with irradiated fuel in the reactor vessel.

Operating Mode Applicability: Cold Shutdown Refueling Emergency Action Levels: (1 or 2)

1. An UNPLANNED event results in RCS temperature exceeding the Technical Specification cold shutdown temperature limit

2. Loss of all RCS temperature and RPV level indication for > 15 minutes.

Differences:

None Deviations:

None

ANO NEI EAL Deviations and Differences Page 17 of 86 COLD SHUTDOWNIREFUELING SYSTEM MALFUNCTION Cu5 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Fuel clad degradation Operating Mode Applicability: Cold Shutdown Refueling Example Emergency Action Levels: (1 or 2)

1. (Site-specific) radiation monitor readings indicating fuel clad degradation greater than Technical Specification allowable limits.

2. (Site-specific) coolant sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits.

Differences:

1. NEI 99-01 Rev. 4 EAL #2 was renumbered EAL #1 for ANO's EALs.

2. ANO does not provide a radiation monitor reading equivalent to NEI 99-01 Rev. 4 EAL #1. ANO uses the letdown radiation monitor (if available) as a qualitative indication of potential fuel clad degradation. Indications on the letdown radiation monitor (if available) are used to prompt plant personnel to take an RCS sample for radiochemistry analysis. The results from the analyses are compared to the IDE and specific activity levels to determine the emergency classification.

Deviations:

None.

ANO NEI EAL Deviations and Differences Page 18 of 86 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU6 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of All Onsite or Offsite Communications Capabilities.

Operating Mode Applicability: Cold Shutdown Refueling Example Emergency Action Levels: (1 or 2)

1. Loss of all (site-specific list) onsite communications capability affecting the ability to perform routine operations.

2. Loss of all (site-specific list) offsite communications capability.

Differences:

None Deviations:

None

ANO NEI EAL Deviations and Differences Page 19 of 86 COLD SHUTDOWNIREFUELING SYSTEM MALFUNCTION CU7 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of Required DC Power for Greater than 15 Minutes.

Operating Mode Applicability: Cold Shutdown Refueling Example Emergency Action Level:

1. a. UNPLANNED Loss of Vital DC power to required DC busses based on (site-specific) bus voltage indications.

AND

b. Failure to restore power to at least one required DC bus within 15 minutes from the time of loss.

Differences:

None Deviations:

None

ANO NEI EAL Deviations and Differences Page 20 of 86 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CU8 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Inadvertent Criticality.

Operating Mode Applicability: Cold Shutdown Refueling Example Emergency Action Levels: (1 or 2)

1. An UNPLANNED extended positive period observed on nuclear instrumentation.

2. An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.

Differences:

1. EAL # 2 of NEI 99-01 Rev. 4 was renumbered EAL # 1in ANO's EALs.

2. EAL #1 of NEI 99-01 Rev. 4 was not used at ANO. ANO does not have a period meter.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 21 of 86 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CAI Initiating Condition -- ALERT Loss of RCS Inventory.

Operating Mode Applicability: Cold Shutdown Example Emergency Action Levels: (1 or 2)

1. Loss of RCS inventory as indicated by RPV level less than {site-specific level}.

(low-low ECCS actuation setpoint) (BWR)

(bottom ID of the RCS loop) (PWR)

2. a. Loss of RCS inventory as indicated by unexplained {site-specific} sump and tank level increase AND

b. RCS level cannot be monitored for > 15 minutes Differences:

1. EAL #2 of NEI 99-01 Rev. 4 was renumbered EAL #1 in ANO's EALs.

2. ANO does not use EAL #1 of NEI 99-01 Rev. 4 that provides for a specific level indication. RVLMS will not monitor level below the bottom ID of the RCS loop.

3. In EAL #1 of ANO's EALs, "reactor vessel inventory" was used in place of "RCS inventory".

Deviations:

None

ANO NEI EAL Deviations and Differences Page 22 of 86 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA2 Initiating Condition -- ALERT Loss of RPV Inventory with Irradiated Fuel in the RPV.

Operating Mode Applicability: Refueling Example Emergency Action Levels: (1 or 2)

1. Loss of RPV inventory as indicated by RPV level less than {site-specific level).

(low-low ECCS actuation setpoint) (BWR)

(bottom IDof the RCS loop) (PWR)

2. a. Loss of RPV inventory as indicated by unexplained {site-specific) sump and tank level increase AND

b. RPV level cannot be monitored for > 15 minutes Differences:

1. EAL #2 of NEI 99-01 Rev. 4 was renumbered EAL #1 in ANO's EALs

2. ANO does not use EAL #1 of NEI 99-01 Rev. 4 that provides for a specific level indication. RVLMS will not monitor level below the bottom ID of the RCS loop.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 23 of 86 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA3 Initiating Condition -- ALERT Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses.

Operating Mode Applicability: Cold Shutdown Refueling Defueled Example Emergency Action Level:

1. a. Loss of power to (site-specific) transformers.

AND

b. Failure of (site-specific) emergency generators to supply power to emergency busses.

AND

c. Failure to restore power to at least one emergency bus within 15 minutes from the time of loss of both offsite and onsite AC power.

Differences:

1. The word 'required" was used in ANO's CA3 in place of "essential" as used in NEI 99-01 Rev. 4 CA3 for ANO terminology.

2. EAL lb was reworded for human factors concerns.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 24 of 86 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CA4 Initiating Condition -- ALERT Inability to Maintain Plant in Cold Shutdown with Irradiated Fuel in the RPV.

Operating Mode Applicability: Cold Shutdown Refueling Example Emergency Action Levels: (EAL 1 or 2 or 3)

1. With CONTAINMENT CLOSURE and RCS integrity not established an UNPLANNED event results in RCS temperature exceeding the Technical Specification cold shutdown temperature limit.

2. With CONTAINMENT CLOSURE established and RCS integrity not established or RCS inventory reduced an UNPLANNED event results in RCS temperature exceeding the Technical Specification cold shutdown temperature limit for greater than 20 minutes.

3. An UNPLANNED event results in RCS temperature exceeding the Technical Specification cold shutdown temperature limit for greater than 60 minutes or results in an RCS pressure increase of greater than {site specific) psig.

Differences:

None Deviations:

None

ANO NEI EAL Deviations and Differences Page 25 of 86 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CS1 Initiating Condition -- SITE AREA EMERGENCY Loss of RPV Inventory Affecting Core Decay Heat Removal Capability.

Operating Mode Applicability: Cold Shutdown Example Emergency Action Levels: (1 or 2)

1. With CONTAINMENT CLOSURE not established:

a. RPV inventory as indicated by RPV level less than {site-specific level}

(6" below the low-low ECCS actuation setpoint)

(BWR)

(6" below the bottom ID of the RCS loop)

(PWR)

OR

b. RPV level cannot be monitored for > 30 minutes with a loss of RPV inventory as indicated by unexplained {site-specific} sump and tank level increase

2. With CONTAINMENT CLOSURE established

a. RPV inventory as indicated by RPV level less than TOAF OR

b. RPV level cannot be monitored for > 30 minutes with a loss of RPV inventory as indicated by either:

Unexplained {site-specific} sump and tank level increase

Erratic Source Range Monitor Indication Differences:

1. ANO added Core Exit Thermocouples to ANO's EAL #2a as another means of monitoring core decay heat removal capabilities.

2. ANO does not use EAL #laor EAL #2a of NEI 99-01 Rev. 4 that provides for a specific level indication. RVLMS will not monitor level below the bottom ID of the RCS loop. For EAL la, a loss of reactor vessel inventory as indicated by various sump and tank level changes was used in place of reactor vessel level indications.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 26 of 86 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CS2 Initiating Condition -- SITE AREA EMERGENCY Loss of RPV Inventory Affecting Core Decay Heat Removal Capability with Irradiated Fuel in the RPV.

Operating Mode Applicability: Refueling Example Emergency Action Levels: (1 or 2)

1. With CONTAINMENT CLOSURE not established:

a. RPV inventory as indicated by RPV level less than {site-specific level}

(6" below the low-low ECCS actuation setpoint) (BWR)

(6" below the bottom IDof the RCS loop) (PWR)

OR

b. RPV level cannot be monitored with Indication of core uncovery as evidenced by one or more of the following:

Containment High Range Radiation Monitor reading > {site-specific}

setpoint

Erratic Source Range Monitor Indication

Other {site-specific} indications

2. With CONTAINMENT CLOSURE established

a. RPV inventory as indicated by RPV level less than TOAF OR

b. RPV level cannot be monitored with Indication of core uncovery as evidenced by one or more of the following:

Containment High Range Radiation Monitor reading > {site-specific}

setpoint

Erratic Source Range Monitor Indication

Other {site-specific} indications Differences:

1. ANO added Core Exit Thermocouples to ANO EAL #2a as another means of monitoring core decay heat removal capabilities.

2. ANO added monitoring of tank and sump levels to ANO EAL #2a as another means of monitoring core decay heat removal capabilities.

ANO NEI EAL Deviations and Differences Page 27 of 86

3. ANO did not use a setpoint for Containment High Range Radiation Monitor in EAL #2b because ANO's Containment High Range Radiation Monitors have not been analyzed for a setpoint that corresponds to core uncovery.

4. ANO does not use EAL #la or EAL #2a of NEI 99-01 Rev. 4 that provides for a specific level indication. RVLMS will not monitor level below the bottom ID of the RCS loop. Various sump and tank and level rises were used as an indication of the loss of reactor vessel inventory.

5. InANO's EAL #1, a loss of reactor vessel inventory in conjunction with the inability to monitor reactor vessel level for greater than 30 minutes was used as the EAL for conditions when containment closure was not established.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 28 of 86 COLD SHUTDOWN/REFUELING SYSTEM MALFUNCTION CGI Initiating Condition -- GENERAL EMERGENCY Loss of RPV Inventory Affecting Fuel Clad Integrity with Containment Challenged with Irradiated Fuel in the RPV.

Operating Mode Applicability: Cold Shutdown Refueling Example Emergency Action Level: (1 and 2 and 3)

1. Loss of RPV inventory as indicated by unexplained (site-specific} sump and tank level increase

2. RPV Level:

a. less than TOAF for > 30 minutes OR

b. cannot be monitored with Indication of core uncovery for > 30 minutes as evidenced by one or more of the following:

Containment High Range Radiation Monitor reading > {site-specific) setpoint

Erratic Source Range Monitor Indication

Other {site-specific) indications

3. {Site specific} indication of CONTAINMENT challenged as indicated by one or more of the following:

Explosive mixture inside containment

Pressure above {site specific) value

CONTAINMENT CLOSURE not established

Secondary Containment radiation monitors above {site specific) value (BWR only)

Differences:

1. ANO did not use a setpoint for Containment High Range Radiation Monitor in EAL #2 b because ANO's Containment High Range Radiation Monitors have not been analyzed for a setpoint that corresponds to core uncovery.

2. ANO added Core Exit Thermocouples to ANO EAL #2b as another means of monitoring for core uncovery.

ANO NEI EAL Deviations and Differences Page 29 of 86

3. ANO does not use EAL #2a of NEI 99-01 Rev. 4 that provides for a specific level indication. RVLMS will not monitor level below the bottom ID of the RCS loop.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 30 of 86 EVENTS RELATED TO ISFSI E-HUI Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Damage to a loaded cask CONFINEMENT BOUNDARY.

Operating Mode Applicability: Not applicable Example Emergency Action Level: (1 or 2 or 3)

1. Natural phenomena events affecting a loaded cask CONFINEMENT BOUNDARY.

(site-specific list)

2. Accident conditions affecting a loaded cask CONFINEMENT BOUNDARY.

(site-specific list)

3. Any condition in the opinion of the Emergency Director that indicates loss of loaded fuel storage cask CONFINEMENT BOUNDARY.

Differences:

None Deviations:

None

ANO NEI EAL Deviations and Differences Page 31 of 86 EVENTS RELATED TO ISFSI E-HU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Confirmed Security Event with potential loss of level of safety of the ISFSI.

Operating Mode Applicability: Not applicable Example Emergency Action Levels:

1. Security Event as determined from (site-specific) Security Plan and reported by the (site-specific) security shift supervision.

Differences:

The word "Commander" was used in EAL #1 instead of "supervision" as used in NEI 99-01 Rev. 4 EAL #1 to be consistent with ANO terminology.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 32 of 86 FUEL CLAD BARRIER EXAMPLE EALS (1 or 2 or 3 or 4 or 5 or 6)

1. Critical Safety Function Status LOSS Core Cooling - Red POTENTIAL LOSS: Core Cooling Orange OR Heat Sink - Red Differences:

1. This EAL was numbered FCB1 in ANO's Fuel Clad Barrier section.

2. ANO-2 does not use Critical Safety Function Status Trees (CSFSTs).

ANO-2 uses Safety Function Status Checks developed by the Combustion Engineering Owner's Group (CEOG) which are based on logic similar to that used for CSFSTs developed for Westinghouse PWRs. However, there is no Safety Function Status Check condition that corresponds directly to Core Cooling - RED path as a loss EAL. Therefore, the loss EAL is incorporated into FCB2 based on Core Exit Thermocouple readings. Similarly, the Potential Loss EAL corresponding to Core Cooling

= Orange is addressed through the FCB2 Potential Loss EAL based on CET readings corresponding to a loss of subcooling. A similar approach was taken for ANO-1, which doesn't use the Critical Safety Function concept in EOPs. This is consistent with the NEI 99-01 basis for the Fuel Clad Barrier CET EALs, which states that they are included for plants which don't have a CSF scheme.

To implement the NEI concern for Heat Sink - Red, indicating an extreme challenge to the Heat Sink Safety Function, the decision to implement Once Through Cooling due to a loss of the SGs as an effective means of removing heat from the RCS was used. The SGs are the preferred means of core heat removal, and Once Through Cooling is the method of last resort for core cooling. As in NEI 99-01, this is considered to be a challenge to both the Fuel Clad Barrier and the RCS Barrier. Used safety function status/functional recovery instead of critical safety function stated for terminology terms for ANO.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 33 of 86 FUEL CLAD BARRIER EXAMPLE EALS

2. Primary Coolant Activity Level LOSS Coolant Activity GREATER THAN (site specific) Value POTENTIAL LOSS: Not Applicable Differences:

This EAL was numbered FCB2 in ANO's Fuel Clad Barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 34 of 86 FUEL CLAD BARRIER EXAMPLE EALS

3. Core Exit Thermocouple Readings LOSS Greater THAN (site specific) degree F POTENTIAL LOSS: Greater THAN (site specific) degree F Differences:

This EAL was numbered FCB3 in ANO's Fuel Clad Barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 35 of 86 FUEL CLAD BARRIER EXAMPLE EALS

4. Reactor Vessel Water Level LOSS Not Applicable POTENTIAL LOSS: Level LESS than (site specific) value Differences:

1. This EAL was numbered FCB4 in ANO's Fuel Clad Barrier section.

2. The Reactor Vessel Level Monitoring Systems at ANO do not provide positive indication of core uncovery. The level indication provided is used to monitor the approach to and recovery from ICC conditions, but the CETs are used to identify core uncovery, and are the only positive indication of core uncovery. Consistent with this approach, RVLMS is used as an indication of potential core uncovery only if CET indication is unavailable. ANO does not use CSFSTs.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 36 of 86 FUEL CLAD BARRIER EXAMPLE EALS

5. Containment Radiation Monitoring LOSS Containment rad monitor reading GREATER THAN (site specific) R/hr POTENTIAL LOSS: Not Applicable Differences:

This EAL was numbered FCB5 in ANO's Fuel Clad Barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 37 of 86 FUEL CLAD BARRIER EXAMPLE EALS

6. Other (Site-Specific) Indications LOSS (Site specific) as applicable POTENTIAL LOSS: (Site specific) as applicable Differences:

1. This EAL was numbered FCB6 in ANO's Fuel Clad Barrier section.

2. ANO used core damage assessment as the "other" indication of fuel clad barrier loss.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 38 of 86 FUEL CLAD BARRIER EXAMPLE EALS

7. Emergency Director Judgment Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the Fuel Clad Barrier Differences:

This EAL was numbered FCB7 in ANO's Fuel Clad Barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 39 of 86 RCS BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6)

1. Critical Safety Function Status LOSS Not Applicable POTENTIAL LOSS: RCS Integrity - Red or Heat Sink- Red Differences:

1. This EAL was numbered RCB1 in ANO's RCS Barrier section.

2. ANO-2 does not use Critical Safety Function Status Trees (CSFSTs).

ANO-2 uses Safety Function Status Checks developed by the Combustion Engineering Owner's Group (CEOG) which are based on logic similar to that used for CSFSTs developed for Westinghouse PWRs. ANO-1 doesn't use the Critical Safety Function concept in its EOPs.

To implement the NEI concern for Heat Sink - Red, indicating an extreme challenge to the Heat Sink Safety Function, the decision to implement Once Through Cooling due to a loss of the SGs as an effective means of removing heat from the RCS was used. The SGs are the preferred means of core heat removal, and Once Through Cooling is the method of last resort for core cooling. As in NEI 99-01, this is considered to be a challenge to both the Fuel Clad Barrier and the RCS Barrier.

If RCS pressure is greater than 2450 PSIG (Unit 1) and 2465 PSIA (Unit

2) and NOT lowering, RCS integrity is challenged, in that it represents a possible uncontrolled overpressurization of the RCS. For ANO-1, the combination of the ERV setpoint, the Pressurizer Code Safety Setpoints, the RCS High Pressure Trip, and the DSS High pressure trip, in conjunction with recovery actions are all expected to be able to lower RCS pressure below 2450 PSIG. For ANO-2, the Pressurizer Code Safety Setpoints, the RCS High Pressure Trip, and the DSS High pressure trip, in conjunction with recovery actions are all expected to be able to lower RCS pressure below 2465 PSIA. Therefore, indications of sustained RCS pressure above 2465 PSIA and not lowering is regarded as a challenge to RCS integrity.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 40 of 86 RCS BARRIER EXAMPLE EALs

2. RCS Leak Rate LOSS GREATER THAN available makeup capacity as indicated by a loss of RCS subcooling POTENTIAL LOSS: Unisolable leak exceeding the capacity of one charging pump in the normal charging mode Differences:

This EAL was numbered RCB2 in ANO's RCS Barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 41 of 86 RCS BARRIER EXAMPLE EALs

3. SG Tube Rupture LOSS SGTR that results in an ECCS (SI) Actuation POTENTIAL LOSS: Not Applicable Differences:

This EAL was numbered RCB3 in ANO's RCS Barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 42 of 86 RCS BARRIER EXAMPLE EALs

4. Containment Radiation Monitoring LOSS Containment rad monitor reading GREATER than (site-specific) R/hr POTENTIAL LOSS: Not Applicable Differences:

This EAL was numbered RCB4 in ANO's RCS barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 43 of 86 RCS BARRIER EXAMPLE EALs

5. Other (Site-Specific) Indications LOSS (Site-specific) as applicable POTENTIAL LOSS: (Site-specific) as applicable Differences:

This EAL was not implemented at ANO because there are no other site-specific indicators available for this EAL.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 44 of 86 RCS BARRIER EXAMPLE EALs

6. Emergency Director Judgment Any condition in the opinion of the Emergency Director that indicate Loss or Potential Loss of the RCS Barrier Differences:

This EAL was numbered RCBS in ANO's RCS Barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 45 of 86 CONTAINMENT BARRIER EXAMPLE EALs: (1 or 2 or 3 or 4 or 5 or 6 or 7 or8)

1. Critical Safety Function Status LOSS Not Applicable POTENTIAL LOSS: Containment - Red Differences:

ANO-2 does not use Critical Safety Function Status Trees (CSFSTs). ANO-2 uses Safety Function Status Checks developed by the Combustion Engineering Owner's Group (CEOG) which are based on logic similar to that used for CSFSTs developed for Westinghouse PWRs. However, there is no Safety Function Status Check condition that corresponds directly to Containment - Red.

ANO-1 doesn't use the Safety Function Status concept in its EOPs. Therefore, this EAL was not used for ANO. The Containment Barrier is adequately addressed in the other Containment Barrier EALs.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 46 of 86 CONTAINMENT BARRIER EXAMPLE EALs

2. Containment Pressure LOSS Rapid unexplained decrease following initial increase OR Containment pressure or sump level not consistent with LOCA conditions POTENTIAL LOSS: Design pressure and increasing hydrogen concentration > 4%

OR Pressure greater than containment depressurization actuation setpoint with less than one full train of depressurization equipment running Differences:

This EAL was numbered CNB1 in ANO's Containment Barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 47 of 86 CONTAINMENT BARRIER EXAMPLE EALs

3. Core Exit Thermocouples LOSS Not Applicable POTENTIAL LOSS: Core exit thermocouples in excess of 1200 degrees and restoration procedures not effective within 15 minutes; or core exit thermocouples in excess of 700 degrees with reactor vessel level below top of active fuel and restoration procedures not effective within 15 minutes Differences:

1. The Reactor Vessel Level Monitoring Systems at ANO do not provide positive indication of core uncovery. The level indication provided is used to monitor the approach to and recovery from ICC conditions, but the CETs are used to identify core uncovery, and are the only positive indication of core uncovery. Consistent with this approach, RVLMS is used as an indication of potential core uncovery only if CET indication is unavailable. Therefore this EAL was written in terms of CET temperatures only. The 700 degrees with reactor vessel level below the top of active fuel does not apply at ANO.

2. This EAL was numbered CNB2 in ANO's Containment Barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 48 of 86 CONTAINMENT BARRIER EXAMPLE EALs

4. SG Secondary Side Release With Primary To Secondary Leakage LOSS RUPTURED S/G is also FAULTED outside of containment OR Primary-to-secondary Leakrate greater than 10 gpm with non-isolable steam release from affected S/G to the environment POTENTIAL LOSS: Not Applicable Differences:

1. This EAL was numbered CNB3 in ANO's Containment Barrier section.

2. ANO did not use first part of the EAL since the two EALs in NEI-99-01 Rev. 4 were considered redundant.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 49 of 86 CONTAINMENT BARRIER EXAMPLE EALs

5. Containment Isolation Valve Status After Containment Isolation LOSS Valve(s) not closed AND downstream pathway to the environment exists POTENTIAL LOSS: Not Applicable Differences:

This EAL was numbered CNB4 in ANO's Containment Barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 50 of 86 CONTAINMENT BARRIER EXAMPLE EALs

6. Significant Radioactive Inventory in Containment LOSS Not Applicable POTENTIAL LOSS: Containment rad monitor reading GREATER THAN (site-specific) R/hr Differences:

This EAL was numbered CNB5 in ANO's Containment Barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 51 of 86 CONTAINMENT BARRIER EXAMPLE EALs

7. Other (Site-Specific) Indications LOSS (Site specific) as applicable POTENTIAL LOSS: (Site-specific) as applicable Differences:

1. This EAL was numbered CNB6 in ANO's Containment Barrier section.

2. ANO used Core Damage assessment as other site specific indications.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 52 of 86 CONTAINMENT BARRIER EXAMPLE EALs

8. Emergency Director Judgment Any condition in the opinion of the Emergency Director that indicates Loss or Potential Loss of the Containment barrier.

Differences:

This EAL was numbered CNB7 in ANO's Containment Barrier section.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 53 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HUI Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Natural and Destructive Phenomena Affecting the PROTECTED AREA.

Operating Mode Applicability: All Example Emergency Action Level: (1 or 2 or 3 or 4 or 5 or 6 or 7)

1. (Site-Specific) method indicates felt earthquake.

2. Report by plant personnel of tornado or high winds greater than (site-specific) mph striking within PROTECTED AREA boundary.

3. Vehicle crash into plant structures or systems within PROTECTED AREA boundary.

4. Report by plant personnel of an unanticipated EXPLOSION within PROTECTED AREA boundary resulting in VISIBLE DAMAGE to permanent structure or equipment.

5. Report of turbine failure resulting in casing penetration or damage to turbine or generator seals.

6. Uncontrolled flooding in (site-specific) areas of the plant that has the potential to affect safety related equipment needed for the current operating mode.

7. (Site-Specific) occurrences affecting the PROTECTED AREA.

Differences:

1. NEI 99-01 Rev. 4 HU1 was renumbered to HU6 in ANO's EALs for formatting purposes.

2. ANO divided EAL #7 of NEI 99-01 Rev. 4 into EAL #7 and EAL #8.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 54 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT FIRE Within PROTECTED AREA Boundary Not Extinguished Within 15 Minutes of Detection.

Operating Mode Applicability: All Example Emergency Action Level:

1. FIRE in buildings or areas contiguous to any of the following (site-specific) areas not extinguished within 15 minutes of control room notification or verification of a control room alarm:

(Site-specific) list Differences:

NEI 99-01 Rev. 4 HU2 was renumbered to HU4 in ANO's EALs for formatting purposes.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 55 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU3 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Release of Toxic or Flammable Gases Deemed Detrimental to Normal Operation of the Plant.

Operating Mode Applicability: All Example Emergency Action Levels: (I or 2)

1. Report or detection of toxic or flammable gases that has or could enter the site area boundary in amounts that can affect NORMAL PLANT OPERATIONS.

2. Report by Local, County or State Officials for evacuation or sheltering of site personnel based on an offsite event.

Differences:

1. NEI 99-01 Rev. 4 HU3 was renumbered to HU5 in ANO's EALs for formatting purposes.

2. In EAL #1, ANO used 'occupied areas of the site" in place of "site area boundary" as used in the NEI EAL.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 56 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU4 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the Plant.

Operating Mode Applicability: All Example Emergency Action Levels:

1. Security events as determined from (site-specific) Safeguards Contingency Plan and reported by the (site-specific) security shift supervision

2. A credible site specific security threat notification.

Differences:

1. NEI 99-01 Rev. 4 HU4 was renumbered to HU1 in ANO's EALs for formatting purposes.

2. ANO used 'Security Shift Commander" instead of "Security shift supervision".

Deviations:

None

ANO NEI EAL Deviations and Differences Page 57 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HU5 Initiating Condition - NOTIFICATION OF UNUSUAL EVENT Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of a NOUE.

Operating Mode Applicability: All Example Emergency Action Level:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring offsite response or monitoring are expected unless further degradation of safety systems occurs.

Differences:

NEI 99-01 Rev. 4 HU5 was renumbered to HU2 in ANO's EALs for formatting purposes.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 58 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HAI Initiating Condition -- ALERT Natural and Destructive Phenomena Affecting the Plant VITAL AREA.

Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2 or 3 or 4 or 5 or 6)

1. (Site-Specific) method indicates Seismic Event greater than Operating Basis Earthquake (OBE).

2. Tornado or high winds greater than (site-specific) mph within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to any of the following plant structures / equipment or Control Room indication of degraded performance of those systems.

Reactor Building

Intake Building

Ultimate Heat Sink

Refueling Water Storage Tank

Diesel Generator Building

Turbine Building

Condensate Storage Tank

Control Room

Other (Site-Specific) Structures.

3. Vehicle crash within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to any of the following plant structures or equipment therein or control indication of degraded performance of those systems:

Reactor Building

Intake Building

Ultimate Heat Sink

Refueling Water Storage Tank

Diesel Generator Building

Turbine Building

Condensate Storage Tank

Control Room

Other (Site-Specific) Structures.

4. Turbine failure-generated missiles result in any VISIBLE DAMAGE to or penetration of any of the following plant areas: (site-specific) list.

ANO NEI EAL Deviations and Differences Page 59 of 86

5. Uncontrolled flooding in (site-specific) areas of the plant that results in degraded safety system performance as indicated in the control room or that creates industrial safety hazards (e.g., electric shock) that precludes access necessary to operate or monitor safety equipment.

6. (Site-Specific) occurrences within PROTECTED AREA boundary and resulting in VISIBLE DAMAGE to plant structures containing equipment necessary for safe shutdown, or has caused damage as evidenced by control room indication of degraded performance of those systems.

Differences:

NEI 99-01 Rev. 4 HA1 was renumbered to HA6 in ANO's EALs for formatting purposes.

ANO's EAL #2 and EAL #3 did not use "Turbine Building" since no vital area is within the Turbine building.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 60 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA2 Initiating Condition -- ALERT FIRE or EXPLOSION Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown.

Operating Mode Applicability: All Example Emergency Action Level:

1. FIRE or EXPLOSION in any of the following (site-specific) areas:

(Site-specific) list AND Affected system parameter indications show degraded performance or plant personnel report VISIBLE DAMAGE to permanent structures or equipment within the specified area.

Differences:

NEI 99-01 Rev. 4 HA2 was renumbered to HA4 in ANO's EALs for formatting purposes.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 61 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA3 Initiating Condition -- ALERT Release of Toxic or Flammable Gases Within or Contiguous to a VITAL AREA Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or Establish or Maintain Safe Shutdown.

Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)

1. Report or detection of toxic gases within or contiguous to a VITAL AREA in concentrations that may result in an atmosphere IMMEDIATELY DANGEROUS TO LIFE AND HEALTH (IDLH).

2. Report or detection of gases in concentration greater than the LOWER FLAMMABILITY LIMIT within or contiguous to a VITAL AREA.

Differences:

NEI 99-01 Rev. 4 HA3 was renumbered to HA5 in ANO's EALs for formatting purposes.

ANO's EALs use the word "adjacent" instead of "contiguous" to fit ANO terminology.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 62 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA4 Initiating Condition -ALERT Confirmed Security Event in a Plant PROTECTED AREA.

Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)

1. INTRUSION into the plant PROTECTED AREA by a HOSTILE FORCE.

2. Other security events as determined from (site-specific) Safeguards Contingency Plan and reported by the (site-specific) security shift supervision Differences:

1. NEI 99-01 Rev. 4 was renumbered to HA1 in ANO's EALs for formatting purposes.

2. ANO's EALs use the word "Commander" instead of "supervision" to fit ANO terminology.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 63 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA5 Initiating Condition -- ALERT Control Room Evacuation Has Been Initiated.

Operating Mode Applicability: All Example Emergency Action Level:

1. Entry into (site-specific) procedure for control room evacuation.

Differences:

1. NEI 99-01 Rev. 4 HA5 was renumbered to HA3 in ANO's EALs for formatting purposes.

2. ANO re-worded the EAL to "Control Room evacuation in progress" since, in some cases, entry into the remote or alternate shutdown procedure may not require the evacuation of the Control Room.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 64 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HA6 Initiating Condition -ALERT Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of an Alert.

Operating Mode Applicability: All Example Emergency Action Level:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or likely potential substantial degradation of the level of safety of the plant. Any releases are expected to be limited to small fractions of the EPA Protective Action Guideline exposure levels.

Differences:

NEI 99-01 Rev. 4 HA6 was renumbered to HA2 in ANO's EALs for formatting purposes.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 65 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS1 Initiating Condition - SITE AREA EMERGENCY Confirmed Security Event in a Plant VITAL AREA.

Operating Mode Applicability: All Example Emergency Action Levels: (1 or 2)

1. INTRUSION into the plant VITAL AREA by a HOSTILE FORCE.

2. Other security events as determined from (site-specific) Safeguards Contingency Plan and reported by the (site-specific) security shift supervision Differences:

ANO used the word "Commander" in EAL #2 instead of "supervision" as used in EAL #2 of NEI 99-01 Rev. 4.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 66 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS2 Initiating Condition - SITE AREA EMERGENCY Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established.

Operating Mode Applicability: All Example Emergency Action Level:

1. Control room evacuation has been initiated.

AND Control of the plant cannot be established per (site-specific) procedure within (site-specific) minutes.

Differences:

NEI 99-01 Rev. 4 HS2 was renumbered to HS3 in ANO's EALs for formatting purposes.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 67 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HS3 Initiating Condition - SITE AREA EMERGENCY Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of Site Area Emergency.

Operating Mode Applicability: All Example Emergency Action Level:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or likely major failures of plant functions needed for 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.

Differences:

NEI 99-01 Rev. 4 HS3 was renumbered to HS2 in ANO's EALs for formatting purposes.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 68 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG1 Initiating Condition - GENERAL EMERGENCY Security Event Resulting in Loss Of Physical Control of the Facility.

Operating Mode Applicability: All Example Emergency Action Level:

1. A HOSTILE FORCE has taken control of plant equipment such that plant personnel are unable to operate equipment required to maintain safety functions.

Differences:

None Deviations:

None

ANO NEI EAL Deviations and Differences Page 69 of 86 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY HG2 Initiating Condition - GENERAL EMERGENCY Other Conditions Existing Which in the Judgment of the Emergency Director Warrant Declaration of General Emergency.

Operating Mode Applicability: All Example Emergency Action Level:

1. Other conditions exist which in the judgment of the Emergency Director indicate that events are in process or have occurred which involve actual or imminent substantial core degradation or melting with potential for loss of containment integrity. Releases can be reasonably expected to exceed EPA Protective Action Guideline exposure levels offsite for more than the immediate site area.

Differences:

None Deviations:

None

ANO NEI EAL Deviations and Differences Page 70 of 86 SYSTEM MALFUNCTION Sul Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Loss of All Offsite Power to essential Busses for Greater Than 15 Minutes.

Operating Mode Applicability: Power Operation (1)

Startup (2)

Hot Standby (3)

Hot Shutdown (4)

Example Emergency Action Level:

1. Loss of power to (site-specific) transformers for greater than 15 minutes.

AND At least (site-specific) emergency generators are supplying power to emergency busses.

Differences:

The word "essential" in NEI 99-01 Rev. 4 was changed in ANO's EALs to "vital" 4.16 KV for ANO terminology and plant design.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 71 of 86 SYSTEM MALFUNCTION SU2 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Inability to Reach Required Shutdown Within Technical Specification Limits.

Operating Mode Applicability: Power Operation (1)

Startup (2)

Hot Standby (3)

Hot Shutdown (4)

Example Emergency Action Level:

1. Plant is not brought to required operating mode within (site-specific) Technical Specifications LCO Action Statement Time.

Differences:

None Deviations:

None

ANO NEI EAL Deviations and Differences Page 72of 86 SYSTEM MALFUNCTION SU3 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of Most or All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:

1. UNPLANNED loss of most or all (site-specific) annunciators or indicators associated with safety systems for greater than 15 minutes.

Differences:

NEI 99-01 Rev. 4 SU3 was reformatted to fit ANO's two different plants into one EAL. The EAL for SU3 was also divided into 2 EALs for human factors concerns because of the two different plants.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 73 of 86 SYSTEM MALFUNCTION SU4 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Fuel Clad Degradation.

Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Levels: (1 or 2)

1. (Site-specific) radiation monitor readings indicating fuel clad degradation greater than Technical Specification allowable limits.

2. (Site-specific) coolant sample activity value indicating fuel clad degradation greater than Technical Specification allowable limits.

Differences:

1. NEI 99-01 Rev. 4 EAL #2 was renumbered EAL #1 in ANO's EALs.

2. ANO does not provide a radiation monitor reading equivalent to NEI 99-01 Rev. 4 EAL #1. ANO uses the letdown radiation monitor as a qualitative indication of potential fuel clad degradation. Indications on the letdown radiation monitor are used to prompt plant personnel to sample the RCS for radiochemistry analysis. The results from the analyses are compared to the IDE and specific activity levels to determine the emergency classification.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 74 of 86 SYSTEM MALFUNCTION SU5 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT RCS Leakage.

Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Levels: (1 or 2)

1. Unidentified or pressure boundary leakage greater than 10 gpm.

2. Identified leakage greater than 25 gpm.

Differences:

None Deviations:

None

ANO NEI EAL Deviations and Differences Page 75 of 86 SYSTEM MALFUNCTION SU6 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT UNPLANNED Loss of All Onsite or Offsite Communications Capabilities.

Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Levels: (1 or 2)

1. Loss of all (site-specific list) onsite communications capability affecting the ability to perform routine operations.

2. Loss of all (site-specific list) offsite communications capability.

Deviations:

None Differences:

None

ANO NEI EAL Deviations and Differences Page 76 of 86 SYSTEM MALFUNCTION SU8 Initiating Condition -- NOTIFICATION OF UNUSUAL EVENT Inadvertent Criticality.

OPERATING MODE APPLICABILITY Hot Standby Hot Shutdown Example Emergency Action Level: (1 or 2)

1. An UNPLANNED extended positive period observed on nuclear instrumentation.

2. An UNPLANNED sustained positive startup rate observed on nuclear instrumentation.

Differences:

1. NEI 99-01 Rev. 4 EAL #2 was renumbered EAL #1 in ANO's EALs.

2. EAL #1 of NEI 99-01 Rev. 4 was not used at ANO. ANO does not have a period meter.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 77 of 86 SYSTEM MALFUNCTION SA2 Initiating Condition -- ALERT Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful.

Operating Mode Applicability: Power Operation Startup Hot Standby Example Emergency Action Level:

1. Indication(s) exist that indicate that reactor protection system setpoint was exceeded and automatic scram did not occur, and a successful manual scram occurred.

Differences:

ANO added wording to SA2 to fit ANO terminology.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 78 of 86 SYSTEM MALFUNCTION SA4 Initiating Condition -- ALERT UNPLANNED Loss of Most or All Safety System Annunciation or Indication in Control Room With Either (1) a SIGNIFICANT TRANSIENT in Progress, or (2) Compensatory Non-Alarming Indicators are Unavailable.

Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:

1. UNPLANNED loss of most or all (site-specific) annunciators or indicators associated with safety systems for greater than 15 minutes.

AND Either of the following: (a or b)

a. A SIGNIFICANT TRANSIENT is in progress.

OR

b. Compensatory non-alarming indications are unavailable.

Differences:

ANO's SA4 uses SPDS and PMS for specified systems that would provide dynamic alarm functions. The word "plant" is used instead of "significant" to fit ANO terminology.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 79 of 86 SYSTEM MALFUNCTION SA5 Initiating Condition -- ALERT AC power capability to essential busses reduced to a single power source for greater than 15 minutes such that any additional single failure would result in station blackout.

Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:

1. AC power capability to site-specific essential busses reduced to a single power source for greater than 15 minutes AND Any additional single failure will result in station blackout.

Differences:

ANO's SA5 used the term "vital" 4.16 KV instead of "essential" as used in NEI 99-01 Rev. 4 SA5.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 80 of 86 SYSTEM MALFUNCTION SS1 Initiating Condition -- SITE AREA EMERGENCY Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses.

Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:

1. Loss of power to (site-specific) transformers.

AND Failure of (site-specific) emergency generators to supply power to emergency busses.

AND Failure to restore power to at least one emergency bus within (site-specific) minutes from the time of loss of both offsite and onsite AC power.

Differences:

1. NEI 99-01 Rev. 4 SS1 was reworded and reformatted for ANO terminology.

2. ANO combined the second and third condition statements from the NEI EAL.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 81 of 86 SYSTEM MALFUNCTION SS2 Initiating Condition -- SITE AREA EMERGENCY Failure of Reactor Protection System Instrumentation to Complete or Initiate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful.

Operating Mode Applicability: Power Operation Startup Example Emergency Action Level:

1. Indication(s) exist that automatic and manual scram were not successful.

Differences:

None Deviations:

None

ANO NEI EAL Deviations and Differences Page 82 of 86 SYSTEM MALFUNCTION SS3 Initiating Condition -- SITE AREA EMERGENCY Loss of All Vital DC Power.

Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:

1. Loss of All Vital DC Power based on (site-specific) bus voltage indications for greater than 15 minutes.

Differences:

None Deviations:

None

ANO NEI EAL Deviations and Differences Page 83 of 86 SYSTEM MALFUNCTION SS4 Initiating Condition -- SITE AREA EMERGENCY Complete Loss of Heat Removal Capability.

Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:

1. Loss of core cooling and heat sink (PWR).

1. Heat Capacity Temperature Limit Curve exceeded (BWR).

Differences:

NEI 99-01 Rev. 4 SS4 was reformatted and reworded for ANO terminology.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 84 of 86 SYSTEM MALFUNCTION SS6 Initiating Condition -- SITE AREA EMERGENCY Inability to Monitor a SIGNIFICANT TRANSIENT in Progress.

Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:

1. a. Loss of most or all (site-specific) annunciators associated with safety systems.

AND

b. Compensatory non-alarming indications are unavailable.

AND

c. Indications needed to monitor (site-specific) safety functions are unavailable.

AND

d. SIGNIFICANT TRANSIENT in progress.

Differences:

1. ANO's EALs use the word "TRANSIENT" which is drawn from ANO documentation and terminology instead of "Significant Transient" as used in NEI 99-01 Rev. 4.

2. SPDS and PMS are specified as systems that would provide dynamic alarm functions.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 85 of 86 SYSTEM MALFUNCTION SGI Initiating Condition -- GENERAL EMERGENCY Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power to Essential Busses.

Operating Mode Applicability: Power Operation Startup Hot Standby Hot Shutdown Example Emergency Action Level:

1. Loss of power to (site-specific) transformers.

AND Failure of (site-specific) emergency diesel generators to supply power to emergency busses.

AND Either of the following: (a or b)

a. Restoration of at least one emergency bus within (site-specific) hours is not likely OR

b. (Site-Specific) Indication of continuing degradation of core cooling based on Fission Product Barrier monitoring.

Differences:

NEI 99-01 Rev. 4 SG1 was reformatted for two different plants.

Deviations:

None

ANO NEI EAL Deviations and Differences Page 86 of 86 SYSTEM MALFUNCTION SG2 Initiating Condition -- GENERAL EMERGENCY Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOT Successful and There is Indication of an Extreme Challenge to the Ability to Cool the Core.

Operating Mode Applicability: Power Operation Startup Example Emergency Action Level:

1. Indications exist that automatic and manual scram were not successful.

AND Either of the following: (a or b)

a. Indication(s) exists that the core cooling is extremely challenged.

OR

b. Indication(s) exists that heat removal is extremely challenged.

Differences:

None Deviations:

None

0CAN020407, Proposed Upgraded Emergency Action Levels

Text

Subject:

Dear Sir or Madam:

3.0 REFERENCES

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GENERAL COMMENT

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