ML022280354

DAEC Emergency Planning Department Procedure Transmittal Acknowledgment Memo

ML022280354
Person / Time
Site:	Monticello, Kewaunee, Point Beach, Prairie Island
Issue date:	08/08/2002
From:	Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML022280354 (102)

Text

Committed to Nuclear Excellence DAEC EMERGENCY PLANNING DEPARTMENT PROCEDURE TRANSMITTAL ACKNOWLEDGEMENT MEMO (TAM-44)

To: NRC-NRR Document Control Desk US NRC Washington DC 20555 Re: EAL Basic Document (Copy 91)

PSM

Title:

n/a Distribution Date: 08108/2002 Effective Date of Change: 0810812002 Return by: 08/22/2002 Please perform the following to your assigned manual. Ifyou have any questions regarding this TAM please contact Don A. Johnson at 319-851-7872.

REMOVE INSERT Rev. 1 Rev. 2 EAL EBD-ORG (PWR: N/R)

Rev. 2 Rev. 3 EAL EBD-S (PWR: 17644)

Rev. 3 Rev. 4 EAL EBD-H (PWR: 17582)

Rev. 3 Rev. 4 EAL EBD-A (PWR: 17385)

Rev. 8 Rev. 9 EAL MASTER (PWR: -)

PERFORMED BY:

Sign Name Date Print Name Please return to: K. Dunlap PSC/Emergency Planning 3313 DAEC Rd.

Palo, IA 52324 To be completedby DAEC EP personnelonly:

Date TAM returned:

EPTools updated: K OLýD

,tALBASES DOCUM-ENT- R 9 INDEX Page 1 of 1 K>

PROCEDURE TITLE REV# REV. DATE Introduction 1 2/01/2000 Definitions 1 2/01/2000 Organization of Basis Information 2 7/12/2002 EBD-A Abnormal Rad Levels/Radiological Effluent Category 3 8/8/2002 I

EBD-F Fission Product Barrier Degradation 3 11/20/2000 Category EBD-H Hazards and Other Conditions Affecting 3 8/8/2002 Plant Safety Category EBD-S System Malfunction Category 2 8/8/2002

s; OR AN ItASIS T ION UINFR PaRev. 2 R I

-2ORGANIZATION OF BASIS INFORMATION Page 1 of 3 EfectiveDate:

The formnat of the EAL Basis information was developecd to address 4raining needs, to facilitate NRC approval, and to facilitate future revisions and 10 CFR 50.54(q) evaluations. ,Each EAL Basis is'organized in the following manner: - '_.;' - . ,

1. Emergency Action Level (EAL) Basis Information Organized by Initiating Condition (IC)

Initiating Condition Identifier For consistency, DAEC has chosen to make its Initiating Condition (IC) identifiers identical to those used in NEI ,document NUMARC/NESP-007. The EAL Technical Basis information is organized-,

by generic IC identifier number and name. 'NUMARC/NESP-007 organized the generic irnformation into four Recognition Categories. These are: , , . .

A - Abnormal Rad Levels/Radiological Effluent .

F - Fission Product Barrier Degradation ,

H - Hazards and Other Conditions Affecting Plant Safety S - System Malfunctions For the A, H, and S recognition categories, all EAL basis information is organized by IC identifier in escalating emergency class order from Unusual Event through General Emergency. For the F recognition category, the initiating conditions are the combinations of fission product barrier losses and ,potential losses that correspond to ,each ..emergency classification level. The individual indicators used on *the fission barrier, table .are separately discussed below. ,..Thle g-enericlC "

identifiers use two letters',followed'by one number. The first letter corresponds to ,the event category as shown above. .The sec6nd letter corres'p`onds t6.jhe emergelncy classification level for the lC :1'. - "_" * ... " "  :

• U - (Notification of) Unusual Event,: . ....

A - Alert S-SiteArea Emergency G - General Emergency The number designates whether the ICtis thie first, second, third, etc., -IC for that recogn'ition category under that ere6rgency.classification. For examplea, SU2,is the designator for the second system Malfunction' recognition 'catego6y IC .in the Unusual Event classification, 'etc. Ge'neric information is used from NEI/NESP-007, Reiisiorn 4, datd May, 1999.

f, tv"I~ 'I ORAIZTBO F SIS- I'NFN*EU,,'- 'Rev. 2 SORGANIZATION OF BASIS INFORMATION.  ;,:,-, . Page 2 of 3 Event Type This is the label of the applicable row for.the EAL Tables shownmin EPIP, Appendix 1. The event type, lists the general area of concern and includes Offsite Rad Conditions,,Onsite Rad Conditions, Natural Disasters, Fire, Other Hazards and Failures, Security, Control Room Evacuation,, EC/OSM Judgment, Loss of Power, RPS Failure, Inability to Maintain Shutdown Conditions,_

Instrumentation/C6mmunication,; Coolant Activity, and Coolant Leak. This-structure was chiosen to be consistent with the previ6uý- EAL preýeýntatio6 which is' already" familiar to' the Ermergency Coordinators and Operations Shift Managers. It is also permissible't6 'organize the generic' information in this manner based .on. the response to- Question, 5.contained in the NUMARC Methodol6gy fo'r Developrmeni of-mei-gehcy"Action L6vels NUMARCINESP-007 R&visio'n 2' Questions and Answers June 1993.

Applicable Operating Modes .. -. , ,,. ,., ...

The 6pplicable operating modes for each Initiating Condition/EmiergencyActionf Lel)is, then'listed" based on NUMARC/NESP-007 mrode! descripti6nrs,. The DAECEALý-i i7s' th6 operating rio6d6s' defined in Technical,_Specifications.Tablel.1-1; These are: '

1 - Run/Power Operation 4 - Cold Sh'utdown(a) '

2- Startup 3 - Hot Shutdowna

.1 1 . - , 5-Refueling(b) ,

(a)AII reactor vessel head closure bolts fully tensioned.,

(b)One or more reactor vessel head closure bolts less than fully tensioned.

Operating' mode applicability of EALs i& based ;on th*, operating mode "that the plant wa's in immediateli b6fore the- event sequence leading to entry intro th emiergenqy classification.`,"For" examp le, events/c6nditionsi addressed bj'EAls applicable' to Run mode'are expected to I1ad' to*'

reactor trip which sh6i)ld bring 'the plant to Hot Shutdown" (M-ode'3). 'Howev-er,' the aplr6priite emergency classification would still be based on the applicable EALs for Run/Power Operation "

(Mode 1) for these events/conditions. If 'ALL" operating modes are specified for the EAL, then the EAL appliesto all modes identified above plus defueled conditions.

EAL Threshold Value The EAL Threshold Value is then listed. This list contains the values, parameters and/or conditions needed fc)r Classification decision making. EAL 'determination is , made from 'the EAL Threshoid Value criteria.'When mor'e than ýorne criteria is pi:6iided,.Iogi "*phrasing is used to, describb whether several conditions need to be met or only one is n6cessary.

- -ALUBASES-DOCDUMENT Rev. 2

-ORGANIZATION OF BASIS IN'FORMATION - '-' Page 3 of 3 DAEC EAL Information This contains the plant-specific information 6sed to implement the genric EALs. This section will also include the basis, 6s appropriate, for deviation from genericEALs§ *As app'opriate, description of any-suppofting balculations, their underlying bases and-+assurnptiort,' and their results are included in this section.

- I *---

R eferences-. .. -. .- - " '. .

The references used to develop'the DAEC EAL Information are listed here; 6s§appropriate.

, +.: ,4" . " ' "- 4

2. Fission Product Barrier Table Indicators The basis information for the fission barrier table indicators is organized similarly to the other basis information described above. For each barrier - fuel clad, RCS, and primary containment - basis information is organized by "Indicator." The indicator is the name forlthe row on the fission' barrier table and is used for donvenieht'"r6tping ofsimila'1-ssymptomsý,-similar-to the "Event Type" used for the A, H, and S EALs described above. zIndicat6*rs "a RPV Level, Leakage, Primary Containment Atmosphere, and EC/OSM Judgment.

After the DAEC Indicator, the applicable generic BWR fissi'n product barrier indicators are then displayed, showing both the generic loss and potential loss conditions, a's 'appIllicable:- Next displayed is the appropriate DAEC information and references. These are displayed in the same manner as the A, H, and S recognition category'basis information desdribed abo{,e. ">

• , + - * . .. . . ,'...-'

-.- '4 . ',+ ,., '- .4 - + + - .,'* A. ,-. " ".

"" S ... . ," -  ;>**4**- *+,'* , IS. + - "

S " 4 ", :.-. ',+ ,

• 4

• --: - . -.. . 4 ,t .. , + *. ,

4 "s -- . - 4 , .\.', ,: c'." , , . * .. i , ,+  :-:'2 ., I 4"

. .. 4.

- +' L .4_. . ... +  : -' , , " . '

• i + " t = + -,

A A EBD-A 7 Rev. 4 I ABNORMALRAD LEVELS/RADIOLOGICAL EFFLUENTC.ATEGORY PAGE 1 of 26 Effective Date:

~I ~)

..' Ie;-

- Date:03--

- Independent Reviewer \

2~~7 PROGEDURE APPROVAL~~&

I am responsible for the technical content of this procedure and for obtaining the necessary approval from the State and County Emergency Management officials prior to implem entatiorn: . . .. -. ::i": -  :

Documentation of State and County Emergency Management approval is via _..

NEP- .2oo6A-03Q- -

Approved by: ManagerEmergenc Date: ,

Manager, Emergency Planning

. TableofContents' ,

AUI - Any Unplanned Release of Gaseous or, Liquid R6adioactivity to the Evivonmiýnt That Ex(ceeds Tw6 Times the Offs'he Dose Assessment Manual (ODAM) and is Expected to Continue For 60 Minutes br Longer .................................................................. ...... A-3 AU2 - Unexpected Increase in Plant Radiation ....... ........................................ A-8 AA1 - Any Unplanned Release of Gaseous or Liquid Radioactivity to the  %

Environment that Exceeds' 200 Times the Offsite, Dose Assessrhent Manual (ODAM) and is Expected to Continue for 15 Minutes or "Long'er."

Loge . ,:. .................. .... :.............. * :.......:.... :............................... ..... A-li...:.'

A -1 1 AA2 . Major Damage to lrradiatedFuel or Loss of Water Level that Has~or Will Result in the Uncovering of Irradiated Fuel Out*ide the/R'eactor Vessel ............................................................................... A-16 AA3 - 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 to Maintain Cold Shutdown., z * . .............. -...................... ............. ..A-1 9

- . ~ 4 ASI - Site Boundary Dose Resulting from an Actual or Imminent Release of.

Gaseous Radioactivity Exceeds 100 rnrem TEDE or 500 rnrem CDDE Thyroid for the Actual or Projected Duration of the Release ............ A-21 AG1 - Sit' Boundary Dose Resulting from an Actual or lmmin-nt Release of I Gaseous Radioactivity that Exceeds 1,000 mrem TEDE or 5,000 mrem CDE Thyroid for the Actuial 6r Projected Du'ationr ofthe Release...-...-,.; ............ ......... ....... .. ................. .............................. A-24

, *4

.4 '.

I

- Rev. 4

, ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT.CATEGORY: PAGE 3'(

AU1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment That Exceeds Two Times the Offsite Dose

' Assessm fMaiiual (ODAM) Limit and is Expectd W Coftifu6 For 60 MMinites or Lo;nger ,.,

r + + * + - * . - , ,

EVENT TYPE: Offsite Rad Conditions 9 , ..+- ., . +' * + . ,, * * . .

OPERATING MODE APPLICABILITY: All.

EAL THRESHOLD VALUE: -(1 or 2 or 3 -or4)+

SAvalid exceed 2X reading on radiation monitors that corresponds to a release that is expected to ODAM level for 60 minutes or longer as ifidicated by:,,

Reactor Building or Turbine Building ventilation (Kaman) rad monitor reading above OR-. -, ,:

"OffghsStack (Kanian) rad monitor reading abo've' 1"E- I pI i '

OR ., . +

LLRPSF (Kaiian) rad monitor reading above 5 E-4 pCi/cc..

OR .

. ahovbmi#3ývP

-GSW

+*tb. 333+3.CS-*

S.. CPS; ' .*,,'

RHRSW & ESW rad monitor reading above 8E3+2 CPS.

OR . . f.'-,;.' ',

RHRSW & ESW 1)ischageana1 rad monitor reading above 1E+3 CPS.

2. Confirmed sample analyses for gaseous or liquid releases indicates 6oncentrations or release rates with a release duration expected to exceed 60 minutes in excess of 2X ODAM limit.

3. Valid perimeter radiation monitor reading of greater than 0.10 mR/hr above normal background for 60 minutes.

4. Valid dose assessment indicating dose rates beyond the site boundary above 0.1 mR/hr TEDE for a period greater than 60 minutes.

K>AUI

i* EBD ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT-CATEGORY 'PAGE 4 of 26 DAEC EAL INFORMATION:

Valid means-that theireading is from. instrumentation' detennined, to. be operable in accordfce with,Ithý Tlinic`al Sp-ecificati6ns or ha- °been verified, bt'ho.e6i', independent

.methods such as indications displayed on the control panels, reports fiom plant personnel, or radiological survey results.

UNPLANNED, as used-in this context, includes any release for;which a radioactivity discharge permit was not prepared; or a release that exceeds, the conditions (&.g, minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. The Emergency Directoi 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 minuies, Also, if an ongoing'release is detected and'the starting iime for that release is unknown, the Emergency Director should, in the* abgence of data to the contrary, assume thatthe release has exceeded 60 minutes - ' .:

The approach faken for calculation of gaseous'radioactive effluent EAL setpoints includes use of the ODAM Table 3-2 source term computed by BWR-GALE for the, DAEC Base Case. The release is assumrie'd to be from a sinole release point. Multiple release points would be difficult to present as explicit EAL threshold values and in any case, are addressed by off-site dose assessment: by MIDAS, which is the preferred method'for determining this condition. The ýal*culation methdds for setpoint determination are from ODAM Section 3.4 and are based bhA Regulatory Guide 1.109 methodology. The table below lists the results of the gaseous effluent- EAL calculations. -Thb Kaman 'extended 'range" capability is used because the Generhl EIectrici Offgas Stack moniitor has a linmited range.ý

- r. -i ,

1AU

l Gaseous EffluentEAs A. *.

Offgas Stk Kaman 9h/0 Turbine Bldg (Kaman 1/2) and

'ReactorBldg (Kam~ri 3/4, 5/6, 7/8)

Maximum flow (CFM) 10,000 72,000 --

Release Limits Concentration Release Rate,' - Concentration Release Rate RL (tCi/cc) Q(Ci/sec) L t(Ci/cc) (ptCi/sec)

Tech Spec

• 1.IE-1 '5.2E+5 -' -'-6.2E 2.1E+4 Unusual Event (2 x TS) 1.20E-1 "' .OE+6 " 1:213-3 '4.2E+4 Alert (60 x TS) -- 6.0E+0' 3.0E+7 "3.7E-2 1.3E+69

__.. .. ____LLRPSF Kman 12 Maximum flow (CFM). 99,000 Release Limits' Concentration . Release Rate Tech Spec - 5.9E-4 . . . 2.8E+4, Unusual Event (2 x TS) 1.OE-3 5.6E+4 Ale' (200 x TS) - -. 0E-1 . . .;.: .. ,:5.6E+6 The off-gas stack .is:treated as ;an -elev~ated release, and-Ahe ,turbine building -and reactor building vents aretr6ated as mixed-mode-ieleases. The!ground level setpoints are taken from -the .defauilt setpoint calculations from :the -quarterly,surveillance tests performed by DAEC Chemistry, technicians.. Reactor Building,. Turbine Building,.LLRPSF -(Low Level Radwaste Processing ,and Storage Facility) and Offgas Stack Noble Gas Monitor alarm setpoints are calculated .based on- achieving the- Tech ýSpec ,instantaneo'us -release 'limit, assuming annual average meteorology as defined in'theODAM: The TechSpec-Liriit:

currently coi-esponds to a' ieactor building'dr turbine building ventilation alarm s6tpoint of 6.2 E-04 ptCi/cc. The monitor alarm setpoint can be periodic~lly adjusted but typically does not vary by much. The DAEC EAL therefore addresses valid radiation levels exceeding 2' times the alarm setpoint for greater than 60 minutes. Rounded off, this corresponds to I E-3 tCi/cc. The corresponding offgas stack monitor value is 1.1E-1 [tCi/cc,'rounded off to 1 E-1 tCi/cc. The Tech Spec Limit currently for the LLRPSF building ventilation alarm setpoint is 5.9 E-04 [tCi/cc. The DAEC EAL'therefore addresses valid radiation levels exceeding 2 times the alarm setpoint for greater than 60 minutes. This corresponds to I E-3 pCi/cc.

Technical specification setpoints for radioactive liquid radiation monitors are 10 times the 10 CFR 20 Appendix B, Table 2, Water Effluent Concentration (WEC) limits. It is the policy of DAEC to process all liquid radwaste so that no release of radioactive liquid to the

- environment is allowed. The radwaste'efflueni line which could be used as a batch release mechanism has a trip function that prevents exceeding the DAEC release limit, however, an, EAL hasl'e'en plrovided.5 The 'trh pathways to'the enviro'nmeni (RHRSW - to cooling-"

t6wver, toer RHRSW - ,6

-- ,t discharge aal) -HS have radiation moni:'*vith

-icb toi- ihra~6g readoutsgoinggt to the AU1

I I [

EABSS CGMETýý EBD-A

, - - ",ý:Rev.

4 ,

i ABNORMAL-RA:LD LE4C/EkS/RADIOLOG ICAL2 EFFLU ENT CA*TEGORY ""PAGE 6 of 26 Control Room: These systefis'6ould bebome contaminated if heat exchanger leaks develop; hl6owever; histonically this' ha not -6ccuirired in the s rvic_,Wate* r i ?ii at DAEC. -These monitors are displaye&d'A panels 1C02 anid ICI1

'Reactor w~ar is the likeli source of contamination tlougi tli ser`vic- water systems' as opposed to floor drain, detergent drain, and chemical waste discharge. The floor drain and

'detergent drains go' to RadwVAste Pf6&ssfing-;fid W-ould be batch released to the Rad'4asfe effluent discharge line (if such ar'elease were to0ocsur)up -The hemical-discharge Su*.i" normally a radioactivity clean' system and is tested by Chemistry toecnsure no conhar ation, prior to dischargingto the canal.:" ... bmt.un t ai The setpoints- for the- three, service" water 'radiation effluent" onitos. because o-'

differences in'detector efficiencies and ba~lkgubl. -setloints based on the sL, reactor water san~pie are listed.llow to show the "differences. -' The r6und&d off readings will be "usedfor the EALs for ease'ofreading the mobitbr scales.- - -. ,..'!

Monitor - TS Limit ." Reading --UE Level -. Alert Level GSW 1,555 CPS 1.5E+3 CPS 3E+3 CPS 3E+5 CPS RHRSW & ESW to cooling '413 CPS 4E+2 CPS l 8E+2 CPS' 81+4 CPS RHRSW & ESW to- -. - 507 CPS 5E+2 CPS IE+3 CPS, 'IE+5 CPS Discharge Canal . .

There are no significant deviations from the generic EALs.: However, DAEC does not have a telemetered radiation monitoring system.. As an alternative, use of field instruments was, considered. It is not practical to establish an EAL, based on' field survey readings of 0.1 mR/hr for greater than 60 minutes because field instruments in use for emergency response, do not have a threshold of detection to meet such criteria.L .

Hourly Whole Body Dose Corresponding to 2 rODAM Limni for Gaseous Release ODAM limit = 500 mrem/year Whole Body D6se 2 x ODAM limit = [2 x 500 mrem/year]/8760 hours/year = 0.114 mrem Whole Body in one

.hour-,

""-1  : " " 'R6undedoffto0.I mieri Do96 -asse ,

ssment g'.MIDAS i. based on the EPA-400 m`ethodology, e.g., use of Total.

Effective Dose Equivalent (TEDE). This is somewhat different fiom whble body dose fromii' gaseous effluents determined by ODAM 'methodology which forims the býasis for the

' AUI

EBD-A ERev. 4 ABOR A R. D LEVE.LS/RADIOLOGICAL E FF.LUENT,' CATEGOR ",:PAGE7 of 26 radiation monitor readings calculated in accordance with the generic. methodology. -The gaseous effluent radiation monitors can only detect' noble gases.,; The -ontribution of iodine's to TEDE could therefore only be determined either by: (1) utilizing MIDAS, or (2) gaseous effluent sampling. DAEC EAL 4 is written in termsr'of TEDE and the' gaseous effluent radiation monitor readings are determined based on ODAM. , . -

REFERENCES:

1. Offsite Dose Assessment Manual.Section 6.12 and 7.1.2 Bases . ,

2. Emergency Plan Implementing" Pr()6eduir (EPIP) 3.3, Dose Assesisment and Protective Action

3. Radiation Protection Calculation No. 95-001-C,,Emergency Actions Levels Based on Effluent Radiation Monitors, Janluiary 24, i995 . .

4. UFSAR Section 11.5, Process ind Effluent Radiation Monitoring and Sampling Systems

5. NEI Methodologyfor Development ofEmergencyAction Levels .NUMARGINESP-007 Revision 4, May 1999 I. C, 44 4. 4 4 4 4 4..- 44 4 444

, , 444 4 I'4 4

.4 44 4 4 4 4' 4444 4

• 44 4.. 14- 44 4 - 3 3, 4 44-44 -, 4

.4 4' 44 444 - 44* -

4 *4'4. 44 44 44 44 4 .4 .4 44 - 4-4 .4 * . 4 4 4 4 4 44444 4 4 4 447

4. 44 4 4 4 4 4 .4 4 4 4 4 44 4. 4 44 4 . 4 4 4 444444 4 44 4 4*44 4 44 44 44 44 4 44, 444 4 .- 44 4 444444 44 .44 4 .444444 4 4 44 444 4 44 4 .4 44444 4 4 44 4 4 4 4 44 . 4 4 4 4 4 4 44 4 44 4 4 4 4 .4 4 4 4 4 .4 4 - 4 . 4 , 4 4 4 444 4.4 4 4 4 44 4 4 4 4 4 44 - 4.444 4 4 4 44 AU1

EA B S S,( C M R EBD-A

-- m& c Rev. 4 .

" *BNORMAL A RAD-LEVEILS/RADIOLOGICL'L EFFLUENT CATEGOR'( "PAGE 8*of 26 AU2 Unexpected Inerease in Plant Radiatii-""

EVENT TYPE: Onsite Rad Conditions OPERATING MODE APPLICABILITY: Al ,

41 .i . .,4 EAL THRESHOLD VALUE: -'(1 or 2 or 3 br 4) 1.rValid WR GEMAC 'Floodiip inzdictdioii (LI24541) coming on scale'in the reactor' refueling cavity witlh all iradikittd fuel assemblies'remaining c6ovýed by water or valid field report to Control Room ofsame. . ,

2. Valid ýfiuel pool level ihdidation (LI-3413) below '36 feet'and loxw'ering with all' irradiated fuel assemblies'remainingt covered b3, water or valid' field report to Control 4Roomofsame:. - . 4 4 4 .. .* ,

3. Valid radiation, reading for'irradiat~d spent fuel in dry storage > .ImR4hr.

4 > 4..

' , '5-.4i

4. Valid., Directt Area tRadiation Monitor readings increasesby a factor of 1000 over

- norm al* levels--p. . ,  :, . . ,,

, I 14 ' " *1i-" " -.

• Normal levels can be considered "asthe lhighest ieading mil the past twent'-foui hours excluding th6'ciifrenftpak value.' ', , " ,, . .

DAEC EAL INFORMATION:

"Thereare no singmficiat deviatibns'froni the'generic EALs: 1lDAEC-doesrnot fi*'ve a spent.

fuel transfer cana1 or oh-sitedry storge of spent fuiel! ". .

Uhic,,trolled'~in6sthai th~e condition is 'not'theresfult of planne'd actioiis by the plant staff in accordance with procedures. Valid means that the reading is from' ismn-stirmentatiIn determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.

There are three methods to determine water level decreases of concern. The first method is by report to the control room. The other methods include use of the Floodup level indicator and the spent fuel pool level indicator. These are further described below.

During preparation for reactor cavity flood up prior to entry into refuel mode, reactor vessel level instrument LI-4541 (WR GEMAC, FLOODUP) on control room panel 1C04 is placed in service by I&C personnel connecting a compensating air signal after the reference leg is disconnected from the reactor head. Normal refuel water level is above the top of the span of this flood up level indicator. A valid indication (e.g., not due to loss of compensating air AU2

EBD-A

.ABNORMAL RAD;LEVELS/RADIOLOG I* ICAL, EF, LUENT-,CA_'IEGORY. ,PAGE Rev. 4 9of 26 signal or other instrument channel .failure) of reactor cavity, level coming on span for this.

instrument is used at DAEC as an indicator of uncontrolled reactor cavity level decrease.

DAEC Technical Specifications require a minimum of 36 feet of water in the spent fuel pool. During refueling, the gates between the reactor cavityand the refueling cavity, are removed and the spent fuel pool level indicator LI 3413 is used to monitor refueling water level. Procedures require that -a normal refueling water level be maintained -at 37. feet 5' inches. A low level alarm actuates when spent fuel pool level drops below 37 -feet 1 inch.

Symptoms of inventory loss at DAEC include visual observationof decreasing waterlevels in reactor cavity or ýspent fuel storage, pool, -'RCator,Bnilding:,(RB) fuel storage. pool radiation monitor or refueling area radiation monitor alarms,- obseivatiori of a decreasing trend on the spent fuel pool water level recorder, and actuation of the spent fuel pool low water level alarm. ,,To eliminate minorlevel perturbations-fromconcern, DAEC uses LI 3413 indicated water level below 36 feet and lowering. ,

"I k1h

~ r b deece 1 I.1 1 Increased radiation revels can be 'detected the'local refueling floor' area radiaftion monitors, the refueling floor- Continuous 'Air Monitor (CAM) 'alarm,- refueling areas radiation monitors, fuel pool ventilation exhaust monitors, and by Standby Gas Treatment (SGBT) System automatic start. Applicable area radiation monitors include those that are displayed on Panel 1C02 and alarmed on Panel 1C041. The DAEC EAL has also been written to reflect the case where an ARM may go offscale high prior to reachIing 1,000 times the normal reading.

NOTE: On Annunciator Panel IC04B, the indicators listed below are expected alarms during pre-planned transfers of highly, radioactive materialthiugh*he affected area. If an HP Technician is present, sending an Operator is not required..,Radiation levels other than those expected should be promptly investigated.' The indicators are 'igh radiation alarms from the Hot Laboratory or Administrative Building, thee neW fuel, storage area, and the radwastebufldmig.;.., ,",,-'. , ,

f -'

- , I I1.

AU2

EBD-A Rev. 4 * .J*

ABNORMALSRAD LtVELS/RADIO0LOG iCAL&EFLU ENT CATFEGORY PAGE 10 of 26 REFEREN/CES: "" * ".. ," .. , .,', -,, . , ,,

1. Alarm Response Procedure (ARP) IC04B, Reactor Water Cleanup and Isolation

2. Technical Specification 3.7.8, Spent Fuel Pool Water Level ,

3. Emergency Plan Implementing Procedure (EPIP) 3.1, Inplant Radiological Monitoring, Attachment 1, ARM Locations . . '

4. Emergency Operating Procedihres (EOP) Basis Document, Breakpoints for RC/L &:L

5. Surveillance Test Procedure (STP) 3.0.0.0-0IPA,;Daily *and Shift Instrmment Checks

6. Integrated Plant Operating Instruction (IPOI) 8, Ota'ge a~nd Refueling Operations ,

7. Fuel & Reactor Component Handling Procedure (F&RCHP) 5; Procedure for Moving Core Components Between Reactor Core and Spent Fuel Pool, Within theý Reactor Coire,,

- I 3S .I.. a 11 1 or Within the Spent Fuel Pool, , I

8. NEI Methodology for Development of Emergency Action, Levels NUMARC/NESP-007 Revision 4, May 1999

-, a

... *a l'

a,~s ~~ ra a a j a~ Z, C;,AU2

S*. Rev. 4 ABNORMAL RAD LEVELS/RADIOLOGICAL-EFFLUENT CATEGORY PAGEI 1 'f 26 AA1 Any Unplanned Release of Gaseous or Liquid Radioactivity to the Environment that Exceeds 200X the Offsite'DoseAssessment Manial (ODAM) Limit and is Expectedito Contifiuiie for 15 Minutes or Longer, EVENT TYPE: OiTsite Rad Conditionsr " - .

OPERATING MODE APPLItABItiTY: All

EAL THRESHOLD VALUE:-" br2'or 3 o04j

'(1 ' ' '"

1.' 'A valid reading on radiation mnbmitors that correspofids 'to 200X ODAM level as indicated by any oftthe following: -'

Reactor Building or Turbine Building ventilation (Kaman) rad monitor reading above 3 E-2 jCi/cc and expected to last for 15 minutes or longer.

Offgas'Stack (Kaman) rfid monitor reading above 6 E+O lCi/cc and expected to last for 15 minutes or longer. - - -

K,_ OR LLRPSF (Kaman) rad monitor reading above I E- I [tCi/cc and expected to last for 15 minutes or longer.

OR GSW rad monitor reading above 3E+5 CPS and expected to last for 15 minutes or longer.

OR RHRSW & ESW rad monitor reading above 8E+4 CPS and expected to last for 15 minutes or longer.

OR RHRSW & ESW Discharge Canal rad monitor reading above 1E+5 CPS and expected to last for 15 minutes or longer.

2. Confirmed sample analyses for gaseous or liquid releases indicates concentrations or release rates with a release duration expected to last for 15 minutes or longer in excess of 200X ODAM limit.

3. Valid site boundary radiation reading of greater than 10 mR/hr above normal background and expected to last for 15 minutes or longer.

4. Valid indication on MIDAS of a release greater than 200X ODAM limit and expected to last for 15 minutes or longer.

SAAA1

RXz EBD;A i

': .... Rev. 4 "-*

I.4.

" ABNORMAL RAD LEV/EES/RADIOLOGICAL EFFILUENT'CATEGORY' " PAGE*I2-of 26 K)j the DAEC EAL INFORMATION:;. ,

Valid means - t he reading is from nstrumentation determined: to be operable' iii accordance with the T-echncal Specificati6ns or, has. been verified b*y'otherindeplendent methods such as indications displayed on the control panels, reports from plant personnel, or, "radiological sui-vey results. Ii a cas6 wheie datý fr6m, Kaman readings is being- used to' determine whetheK -n EAL threshold -value ha*s been exceeded, Valid,means that flow through the associated Kamarl Monitor has been veiiffed-,and does exist as indicated in uCi/sec on SPRAD. . . -,

UNPLANNED,, as, used in this context, includes any. release forl which a radioactivity discharge permit,' was 'not preji~ired,' or a release that exceeds -te conditions .(e.g.,

minimum diluti6oi fl6w, maximum discharge flow, alarm set oi'nts, etc.) on th'e applicable permit. The Emergency Director'should not Nrait until- 15 minutes hag elap§ed, b,-ti,shouild declare the event as soon-as it is determined that the'release duration 'has oWr-wll- likely exceed 15 minutbeL'Also, if an ongoing release- is detected and the starting'tiime for that release isunknown; the Emergency Director -should, in the absence of data to the 6ontrari, ýssufine that the'release lias exceeded 15 minutes' 2-'

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-.ABNORMAL RAD LEVELS/RADIOLOGICAL EFFLUENT, CATEGORY, PAGE. 13 of 26 Gaseous Effluent EALs "Oo ,TurbineBldg(Kaman 1/2)and ffgas Stack. Kaa..9/10*-Reactor Bldg (Kaman 3/4,5/6,7/8)

Maximum flow (CFM) - 10,000 . ,. -'&72,000

• Releasetimits) Concentration', Release Rate Concenitration - Release Rate

.- . ,, , (ItCi/cc): , (Ci/se)'

t , (pCilcc) .'; (pCi/sec Tech Spec .. 1.IE-1, . 5.2E+5 --6.2E:4 "2E+4 Unusual Event (2 x TS) 2.04El '.  :: I10E--" 1.2E-3 4:2E+4 Alert (60 x TS) 6.OE+0 '3.0E+7 1" -3.7E1-2 1 - 1.3E+6 LLRPSF Kaman 12. . .

Maximum flow (CFM) 99,000 Release Limits Concentnition Release Rate

- (.t~Wi/cc) (~/e

.Tech Spec 5.9E-4 2.8E+4' UnustialEvent(2xTS) j 1.0E-3 . , . . - 5.6Et4..

Alert(200xTS) - .01E-1 ... ,- ,. -:.,-5"6E+6... .

The off-gas stack is treated as an elevated release and ghe turbine building and reactor, building vents are treated as mixed-mode releases. The ground level setpoints are taken

-from the default setpoint calculations'from the quarterly surveillance tests performed by DAEC Chrmistiy technicians. Reactor BuildingTurbine Building,-LLRPSF (Low Level

-Radwaste Processing and Storage Facility) and Qffgas Stack Noble Gas Monitor alarm setpoints are calculated based bn achieving the Tech Spec instantaneous release limit

,assuiriing annual average nmeteorilogy -as def -e-d i, -the-ODAM. Th-T~h S13c-Limit currefitly corresponds to.a reactor building or turbine building ventilation alarm setpioint of:

6.2 E-4 pCi/cc. The monitor alarm setpoint can be periodically adjusted but typically does not vary by much. For the Offgas Stack,'Reactor Building and 'Turbine building KAMAN monitor readings, DAEC chose to multiply the technical specification concentration by a factor of 60 (instead of 200) in order to allow for a logical step progression ini monitor.

__setpoints from the AU1 through AA1 to ASI. The DAEC EAL therefore addresses yalid

. radiation levels exceeding 60 times the alarm setpoint for greater than 15 minutes. Rounded down, this corresponds to 3 E-2 pICi/cc. The corresponding offgas stack monitor value is 6.6 pCi/cc, rounded down to 6 E+0 pCi/cc. The -Tech Spec Limit currently for the LLRPSF building ventilation alarm setpoint is 5.9 E-04 pCi/cc. The DAEC EAL therefore addresses

-valid radiation levels exceeding 200 times the alarm setpoint for greater than 15 minutes.

This corresponds to I E-l ItCi/cc.

-Technical specification setpbints for radioactive liquid radiation monitors are -10 times the 10 CFR 20 Appendix B, Table 2, Water Effluent Concentration (WEC) limits. It is the policy of DAEC to process all liquid radwaste so that no release of radioactive liquid to the AA1

[ L(AE-0UE EBD-A:,

,ABNORMAILRAD LCEVELS/FADIOL*OGllCAL EFFLUENTi CATEGORY- 'PFAGE 14 of 26 environment is allowed: The, radwaste effluent line which couild be used as a batch release mechanism haYa , -triplfunction that .prevents exceedingth e DAEC release limit, and therefore no EAL limits are provided. The other pathvWays to the environment (RHRSW - to' cooling tower, RHRSW - to discharge canal) have radiation m0-otors-fvi.'th reiid6its going to the Control' Room. These systems could bec'ome, contaminated if heat ekxcl'inger leaks develop; however, historically this has not occurred in'the service Nater systeffis at DAEC.

,These monitors are displayed on oanels 1C02'and ICIO..

Reactor water is the likely source of contamination through the service water systems as opposed to floor drain,- detergent diain, and chermical Waste discharge. -The floor drain and detergent. drains go, to Radwaste Processing and would be' batchlreleased,'to the Radwaste effluent discharge line (if such'a release were to occur): The chemical'discharge sump is normally a radioactivity clean system and is tested'by Chemistry to ensure no contamination prior to discharging to the canal.- *,  :. .  ;.

The setpoints for the three service water radiation- effluent monitors-vai'y becaise -of differences- in, detector efficiencies', and- background: "Setoints b!ed on the:same jieactor water sample are listed below to show the differences. The rounded off readings Will be used f6r the EAIs for ease of reading the mni6itor scales.'  :* - "

Monitor TS Limit. ' Reading .UE Level Alert Level GSW 1,555 CPS' 1.5E+3 CPS 3E+3 CPS 3E+5 CPS RHRSW & ESW to cooling 413 CPS' 4E+2 CPS 8E+2 CPS 8E+4 CPS tower "

RHRSW &ESWto - 507 CPS 5E+2 CPS 1E+3 CPS 1E+5 CPS Discharge Canal .

DAEC does not have a telemetered radiation monitoring system. As an alternative, DAEC uses ,valid field survey readings outside the site boundary greater than 10 mR/hr or greater than 50 mR/hr CDE Thyroid., ,

Hourly Whole Body Dose CorreSponding to 200 x ODAM Limit for Gaseous Release

" " ODAM limit = 500 mrem/yeari Whole Body 200 x ODAM limit = [200 x 500 mrem/year]/8760 hours/year = 11.4 mrem Whole Body in one hour Rounded off to 10 mrem Dose assessment using MIDAS is, based on the EPA-400 methodology, e.g., use of Total Effective Dose Equivalent (TEDE). This is somewhat different from whole body dose from AAI

) EBD-0 M Rev. 4 "ABNORMAL RAD.LEVELSIRADIOLOGICALEFFLUENT CATEGORY,- PAGE ,15 of 26 gaseous effluents determined, by, ODAM 'methodology: which 'forins, the -basis for'the radiation monitor.readings calculated in AUl in accordafice with the generic ,methodology.

The gaseous e'ffluent radiation-monit6rs can bl~y detectinioble gise'.'`The 66ntribution of iodine's to TEDE could therefore only be determinedeithernby. (1) utilfzing MIDAS, or (2) gaseous effluent sampling. DAEC EAL 4 is~written in terms of TEDE and the ,gaseous effluent radiation monitor readings are determined based on V ,M.

St f t

-, .1

'-4

REFERENCES:

1. Offsite Dose Assessment Manual Section 6.1.2and 7.l.2,Bases" "' V

2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective

'Action " * '

3. Radiation Protection Calculation No. 95-001-C, Ermerlgency Actions Levels Based on Effluent Radiation Monitors, January 24,1995 " ..

4. UFSAR Section 11.5; Process 'and -Effluent Radiation Mohitoring and, Sampling S y ste ms , . , ,. . , . ' '. - J-, , . , -",, , , , ,

• 5., EPA 400-R-92-001, Manual ofProtectiveActiouiGiddesiandProtectiveActi6ns for NuclearIncidents ,.., , L ,, ,, ....

6. NEI Methodology for Development of Emergency Action Levels 'NUMARC/NESP-007 Revision 4, May 1999 .

4'_

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K)j AA2" Majoi Damage'to' Irradiated Fuel or Loss Of.Water'Level that Has "orWill ResuIft in the Uncoverinig of irradiated Fuel' Ouside thie Reactor Vessel ' -,. ,.

EVENT TYPE- OnsitejdR oniitions-OPERATING MODE APPLICABIITY: All EVAL THRESilOLDI VALUJE:-, (1 or2or 3 or 4)

1. Any-of the folloý,ing 'valid. radiation monitor readigs, for the refuel floor area,, fuel handling area, and the fuel bridge'area: ....

"* ARM HI RAD alarm for the Refueling Floor North End, Refueling.Floor South End, New Fuel Storage Area, or Spent Fuel Storage Area:,";,. .

"* Refueling Floor North End, Refueling Floor South End,' orN6)v Fuel Storage'Area'

, ARM-ReadingboqyejlOmR/hr: ., r '

• Spent Fuel Storage Arda ARM Reading above 100 mR .hr'

2. Report of Visual observation of irradiated fuel uncovered.

3, Water' level, reading below 450". a*sindicated 6n LI4541 - (floodup). for the Reactbr

'Refueling Cavity that Will result in Irradiated Fuel becomihg uncovered ornvalid field

'report

, to Control Room of same. , .

4. Valid Spent'Ffil Pool 61water level M` dication (LI:3413) below 16 feet Water Level that will i6sult in lra'-diated Fuel being unciovering or 'valid'field'rep6rt to Control Room of same., .

DAEC EAL INFORMATION:

Valid means that the reading is from instrumentation determined to. be operable in accordance with, the:.Technical

" "- -: -rt *'* Specifications

' **

• ii.or "Ihas

-. been'veniffi6eby

.; ",' ."-. *'oi , 'oith'e

independent p~i , nel r methods such as indicationis displayed on the control p ffbm plant personnel, or radiological survey results. Valid alarms are solely due to damage to irradiated fuel or loss of water level that Airs or will result in the'uncovering of iadiated fýeL ' ,

There are no significant-deviations from the generic EALsU ,'Inrae radtm levels-1nc'ea's-d'ialiation levlsca can bbe detected by the local radiation monitors, in-plant radiological surve , new'fu'el and spent fuel storage area Iadiatiohrnmonitor alarms'displayed on panel 1C04B, fifel 1ool ventilation exhaust monitors, and by Standby 'Gas' Treatneint (SBGT) Systemi automatic sthrt.

Applicable area radiation monitors include RT 9163, RT 9164, RT 9153, and RT 9178.

AA2

ý'L S Eý ' C"

,U E VE B D-A S* . Rev. 4 .

ABNORMAL RAD LEVELS/RADIOLOGICAL'EFFLUENT CATEGORY, 'PAGE 17 of 26 These monitors are located in the north end of the refuel flo6r, 'The. south end of the refuel floor, the new fuel vault area, and near the spent fuel pool; respectively,.

Per ARP IC04B, the applicable area radiation monitor alarms actuatew hen radiation' levels increase above 100 mR/hr in the spent fuel pool area or above 10 mR/hr in the'other three" areas of concern. If a valid actuation of these alarms were to occur, the refueling floor would be~immediately evacuated. Thus, a-report of a-fuel handling accident with either' valid actuation of the fuel area alarms on panel 1C04B ori with measufed iadliatioA levels in-,

the spent fuel pool or north fuel area are used to address the generic concern consistent with:

DAEC design and procedures., ' .

During preparation for reactor cavity flood up prior'to enti'y into refiel mode; reactor vessel level instrument LI-4541 (WR, GEMAC; FLOODUP) on control room"panel 1C04 is placed in service by I&C personnel connecting, a compensating air signal afte# the ieference leg is' disconnected from the reactoi head. Normal refuel water level is'above the top of the span of this ^flood up' level indicator. ,'A valid, bn-scale -indication '(e.g,, not :due -to loss of compensating air signal or other instrument, channel failure) from thisgistnirnrni can be used to determine uncontrolled loss ofwater~level in the-reactor cavity. : ... ,

During refueling, thegates between the reactor caw4ty and the refueling aje removed

.avit" and the spent fuel pool level indicator LI 3413 is used to monitor refueling water level. This.

measures the common water level -inthe reactor cavity and the fuel pool., The bottom of the fuel transfer slot between. the spent fuel pool and the reactor-cavity is, 16, feet above the' bottom of the spent fuel pool. The top of the active fuel in the spent fuel storage racks is slightly, less than 13,,feet 9 inches above 'the' bottom of thi spent fuel pool.,Therefore, postulated failures which drain the reactor cavity through the reactor vessel cannot uncover fuel in the spent fuel storage racks. However, valid indication of spent fuel pool level less than 16 feet would indicate that spent fuel in the storage racks ,may potentially become uncovered.

RFP404 re~utres that Uipon a loss 'ofwater level sittition, that the refueling crew on the refuelingfl6or shall discharge any iel assembly onitlefuel grappls afollows:

• If afuel asse6ifblyis currentlyr being withdrawn'from a slot n the coe6 r spent fuel pool, immediately reinserit Iifito that'slot. - .' ,

0 If a fuel assembly isbeing transferred and is still over or near the core, insert it into the closest available sokt mithe core.., ' "

If a fuel assembyj is ,being transferred and -is over or near, the spent fuel poo, insert it into the closest available slot in the spent fuel racks.,

Following these actions, the refueling floor is to be evacuated of all personnel. The DAEC EAL is written to address the generic concern that a spent fuel assembly was not fully AA2

I I I S~EBD-A,

~~Rev. 4i '

- -ABNORMAL RADLEVELS/kADIOIXdGJCAL EFFLUENT'*C:/*tEGORY " PAIjE 18 covered, by :water. ,',This can-, ither be by visual observati61n of an un6vered, spent ' fel assembly or by-trending'Afel pool level, in the, control room if i spenit fuel assembly could not be placed in a safe storage location specified by F&RCHP'5 as described above.

REFERENCES:

1. Alarm Response Procedure (ARP) 1C04B; Reactor Water Cleanup and Isolation 2., Technical Specification 3.7.8, Spent Fuel Pool Water Level  ; ,

3. Emergency Operating Procedures (EOP) Basis Document, Breakpoints for RC/IL & L'

4. Emergency Plan Implementing Procedure (EPIP) 3;I,pInplant Radiological Monitoring, Attachment 1, ARM Locations

5. Surveillance Test Procedure (STP) 3.0.0,0-01, Daily and Shift Instrument Checks 6.. Integrated Plant Operating Instruction (IPOI) 8, Outage and Refueling Operations

7. Fuel & Reactor Component Handling Procedure RFP404; Procedure for Moving Core I

.Components Betweeri Reactor Core and Spent ,Fuel Pool,',,Within the Reactor- Corey,or-Within the Spefit Fuel Pool, . ' - - ., .,

8. Bechtel Drawing C-492, Reactor Building - Reactor Well, Spent Fuel & Dryer-Separator Pool General Arrangement, Rev. 6 '%,: " .;" .. -.. . I -.

9. Bechtel Drawing C-493, Reactor Building - Spent Fuel Liner Plan Elevations and "Details,Sheet 1, Rev, 6 .,

10' Holtec Iixtemational Drawing No. 1045, Rack Constructio Spent Fuel Storage Ra&s, Rev.33 .  : . ,: ....

11. NEI Methodology for Development of Emergency Action Levels.NUMARC/NESP-007 Revision 4,;May 1999

• . r2

ýbS8D0_Q M`ýNT EBD-A S~Re',. 4 ABNORMAL RAD LEVELS/RADIOLOGICALEFFLUENT CATEGORY' PAGE 19,of 26 AA3 Release of Radioactive Material or Increases in Radiation Levels, Within the Facility That Impedes Operation ofSystems Required to Maintain'Safe Operations or to Establish ort6 Maintain Cold Shutdowin '

EVENT TYPE: Onsite Rad Conditios, ..

OPERATING MODE APPLICA'BILITY: All,*"

EAL THRESHOLD VALUE:or '(1o2j '" "

1. yalid area rad m6nito# (RE9162)' reading GREATER THAN 15;mRhr in -the Control Room. l

2. VýIi'ýdar'eaiad mhonitor'(RE9168) reading GREATER THAN 500 i0Ror at the Remote Shutdown Panel, 1C388.

DAEC EAL INFORMATION:

Valid means that the reading is from instrumentation determined' to be operable in ac6rdahce with the Teclinical Specificationi or has been verified by other indepefident methods such as indications displayed on the control panels, reports from plant personnel, or radiological surveý iesults. "

There are no significant deviations from the generic EALs. Per the UFSAR, the control room is the only area that is required to be continuously occupied to achieve and maintain safe shutdown following design basis accidents. The capability exists for plant shutdown from outside the main control room in the event that the control room becomes uninhabitable using remote shutdown panel 1C388. The RB 757 CRD North ARM-9168 is in the vicinity of the Remote Shutdown Panel and is used to monitor radiation levels to determine habitability for that area.

The EC/OSSshould determine the cause of the increasein radiationlevels and review other EALsfor applicability. Expected increases in monitor readings due to controlled evolutions (such as lifting the steam dryer during refueling) do not result in emergency declaration.

Nor should momentary increases due to events such as resin transfers or controlled movement of radioactive sources result in emergency declaration. In-plant radiation level increases that would result in emergency declaration, are also unplanned, e.g., outside the limits established by an existing radioactive discharge permit.

AA3

.Ai Rev. 4 S,. ABNORM'AL PAD:LEVELS/RADIOLOGICAE EFFLUENT CATEGORY' PAGE 20 of 26

REFERENCES:

1. Alarm Response Procedure (ARP) 1C04B, Reactor-WatereClanup and solation

2. Abnormal Operating Procedure (AOP) 913, Fire

3. Abnormal Operating Procedure (AOP) 914, Security .

4. Abnormal Operating Procedure (AOP) 915, Shutdo'wn OutsideConttol Room

5. Surveillance Test Procedure (STP) 3.0.0.0701, Daily and Shift Instrument Checks

6. Integrated Plant Operating Instruction (IPOI) 8,'Outage and Refuefirg Operations

7. Emergency Plan Implementing Procedure (EPIP) 3.1, Inplant Radiological Monitoring

8. UFSAR Section 6.4, Habitability Systems

9. Bechtel Calculation DA-4, Project Number 265-002, Control Room Habitability, 9/3/80

10. NEI MetAodolj fbi-Dk!qment of Einerg-eniicy-Actio'n Levels' NUMARC/NESP-007 Revision 4, May 1999

- . t , A

• I.

-C Cr - . . - . ..- t -

", ES*OC EBD-A

~Rev. 4' ABNORMAL RADLEVELS/RADIOLOGICAL EFFLUENT-CATEGORY' PAGE'21of 26 AS1 Site Boundary Dose Resulting from an Actual'oi" Imminent Release of Gaseous Radioactivity Exceeds 100 mrem TEDE'o" 500 mrem CDE Thyroid for the Actual or Projected Duration of the Relea'se- ',.

EVEN:TOffsite Rad Conditions, OPERATING ,MODEAPPLICABI1iTY::AAli, ll'E . .. ., -,, .... . . .

EAL THRESHOLDVALUE: ( or:r~~or 2or3), :  ; r, "- *- ,; . .

1. A valid radiation monito'r. readingwhich corresponds, to an offsite dose o 100 harem or Bhbidi 500 mRieatr as indicated by.*e, following: .- ,,.

Reactor Building or Turbine Building ventilation (Kaman) rad monitor reading above 6 E-2 tCi/cc for more than 15 minutes. (Dose assessment not available)

OR Offgas Stack (Kaman) iad monitor readingI above 4 E+ Ci/cc Ifor more than 5 "iminutes. (Dose assessment'not available)

>2. Valid MIDAS dose assessment projection indicates dose consequences greater than 100 mrem TEDE or,500 mrem CDE thyroid.

3. Field survey results indicate site boundary dose rates exceeding 100 mrem/hr expected to continue -for more than one hour; or analyses of field survey samples indicate CDE thyroid of 500 nirem for one hour of inhalation.

DAEC EAL INFORMATION: .,

Valid means that the reading is fr6m instrumenitationr determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiol6gical survey results. In a case where data from Kaman readings is being used to determine whether an EAL threshold value has been exceeded, Valid means that flow through'the associated Kamar Monito has be~n-Verifned and does-exist as indicated in uCi/sec on SPRAD. . , "- ,

The preferred method for declýraiion of AS l'is by riieans of Dose Assessment using the MIDAS computer'model. How-ever, if Kinan monitor ?readings ýre sustained 'for loniger than 15 minutes'and the iecluired MIDAS dose assessments cannot be completed within this period, then the declaration 'can be made using: Kaman readings PROVIDED the readings ASI

EBD-A ABNORMAL RAD LEVELSIRADIOLOGICAL EFFLUENT CATEGORY' Rev. 4 PAGE 22 of 26 are not from an isolated flow path. If Kaman readings are not valid, field survey results miy be utilized: *. ._

DAEC's Meteorological Information and,,DoDse Assessmient Syjstein (MIDAS) was utilized to determine the KAMAN monitor limits. Eiglit separate combinations of release point, source term, meteorological, conditions and, equipment status were analyzed. - Pathways considered were the offgas stack, the turbine building exhaust vent a'nd a single *reactor' building exhaust vent. Multiple release points were not considered., In this same vein, it was assumed that' ohly one of the three reactor building vents is on durfig the release.

The source terms used have been pre-loaded into MIDAS, - and are" the default' mixes associated with a loss of coolant accident (LOCA) and a control rod drop (CRD). The LOCA mix was used in conjunction with'a release via, the offgas 'sfiack vAhile the CRD mix was used for releases via the turbine or reactor briiding veynts. The sourc6e term for a release via the offgas stack is' further impacted by the'status of the standby gas treatment system.

The status of that system was also takefn into consideratiort.,, .-- ,, ,, . - 't

.3 " o .' -. .

Based on 1995 date (NG-.96-0987), the atmospheric stability was classified as Pascal E 33%

of the time.. Consequently,, both classifications were evaluated. Based on the same report, the'most common wind speeds were: '

• I ,

Pascal Class Altitude Speed (mph')*

D 156' .. ... 8-712 D  : 33' 8-12 ,,".,

E 156' - 8-12 E -'"' ' 33' 4-7 Though the temperature setting has no impact on the MIDAS calculations, a value must be entered in order for the pr6gram to run.' Consequently, the,temperature was arbitrarily set at 50 F.', - ..

The rai estimate was set at zetoi to' elin~iin'te 'nydon site washout of radi6active material.,

For the first MIDAS runs a 1Ci/cC concentration was assumed:. Thiel results of these'runs" w6re thefn normalized io ihe limits, thus generating a theoretical- KAM.AN limit. Additional, MIDAS runs were made with these theoretical limits as input to verify the normalization process.In additi6ntfo"the total integrated dose, MIDAS calculates a peak Whole body DDE rate resulting from the plume and a peak thyroid CDE'rate' resulting ,rom inhalation.

Because the 'ASIand AGI KAMAN limits are to be based on ao'bne hour exposure,`

establishing 'concentration, limiti so these peak valuei match, the NUMARC limits is acceptable.

ASI

EBD-A, In 4%. Rev. 4 ABNORMAL RAD LEVELS/RADIOLOGICALEFFLUENT. CATEGORY PAGE 23 of 26

-, Site Area '., General.

Initiating Condition Emergency  % Emergency

_- _ _ _ _ " _ _ __- 'ASI - AGI Valid Turbine'or Reactor Building ventilation rad  ;"" -

monitor (KAMAN) reading for more than 15, 0.06 pCilcc 00, Ci/cc minutes above: lift.. -,

""-* "i . - , , 1. - 2 ,

DAEC does not have a telemetered radiation monitoring system,.- A's anniltemative, DAEC' uses valid field survey readings outside the site boundary to determine if doses are greater than 100 mR/br TEDE or greater than 500 mR/hriCDE Thyrbid.

Dose assessment using MIDAS is based'on the EPA-400 methodology; e.g.; use of Total Effective* Dose, Equivalent (TEDE) and 'Coinmitt~d Dose-Equivalent (CDE) Thyroid.

TEDE-is somewhat different from,whole body dose from gaseous effluents determined by ODAM methodology which forms the basis for the radiation monitor readings calculated in AU1. These factors can introduce differences that are at least as large as those introduced by using TEDE versuswhole bodydose. The gaseous effluent radiationmonitors can only;,.

detect noble gases.; Thecontribution'of iodine's to TEDE and CDE Thýroid could therefore only be determined either by: (1) utilizing the source term -mixture in MIDAS, or (2) gaseous effluent sampling. Therefore, DAEC EAL Threshold Value 4 is wi'ttei in teris of TEDE and CDE Thyroid. ,

REFERENCES:

1. Offsite Dose Assessment Manual, Section 6.1.2 and 7.1.2, Bases

2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective, A ction"- ' * ' " ."- . I . ' . .'- , - .. '

3) Radiition Protection CalculationfNo..95-9001C, Emergency Actions Levels Based on Effluent Radiation Monitors, Janiary 24, 1995

4. Radiation Engineering Calculation No: 96-007-A, Determination of DAEC Radioactive

,Release .Initiating Conditions forAS L& AGI Emergenicy.Classifications, July 3, 1996 5...UFSAR Section 11.5, Process and, Effluent Radiation Monitoring and Sampling

,6., EPA 400-R-92-00l1,'Manual of Protective,-Action .Guides-and Protective Actions for NuclearIntidents":

17. NEI M~thbdologyfor Development .ffEmergency Action Levels NUMARC/NESP-O07.

Revision 4,-Mayi.9 .9 ,

, * < I,. -#"" . .... t

-J ASI

6ý Rev. 4 ý.

ABNORMAL, RADS~~EBD-A':

LEVELS/RADIOIOGICAf EFFLUENT CATEGORY PAGE 24 of 26 -J AGI Site BoundaryDose Resulting from an Actual orImminent Release of Gaseoii Radioactivity that Excee or 5,000 'irem'CDE Thyroid for the Actual 6r'Prsjected DuratioTnE of the Release ,

EVENT TYPE: OflsitelaRd Conditi ons OPERATING ODE A PPLICABILITY, All EAL THRESHOLD VALUE: (1 or2 or 3),' .

1. A valid radiation monitor reading which corresp'onds to an offsite dose of 1000 mrem or 5000 mrem Thyroid as indicated by the following:

Reactor Building or Turbine Building ventilation (Kaman) rad 'moiiitor " rendig"aabove I....

6 E-1 ICi/cc for more than.15minutes. (Dose assessment iiotAVailable)' ' "

OR Offgas Stack (Kaman) rad,monitor reading above, 4 E+2 ptCi/cc for more than 15 minutes. (Dose assessmefit not available).

2. Valid MIDAS dose, assessment projection indicates dose consequences greater than 1,000 mrem TEDE or,5,000 mrem CDE thyroid.

3. Field survey results indicate site boundary dose rates exceeding 1,000 mrem/hr expected to continue for more than one hour; or analyses of field survey samples'indicate CDE thyroid of 5,000 mrem for one hour 6f inhalation.

DAEC EAL INFORMATION:

Validieans,tha-t' the"'readiiig is ff6in inr tnentation- deteimiied' to" be operable' in accordance with the Technical Specifications 'or has been',verified by'othdr independent methfods such as indications' disPlayed on the c6ntrol panels', reports ftilh plant piersohnel, or' radiological survey rfesiilis'.ý In 16a-se"',where

'da't.froii Kainýda' r6adl'mis'm' s o determine' whethex'an EAL *thI*rdsold 'V",iu" fia' bei'ý exceeded,' Vci lid means that fow through the associated Kaman Monitor has been verified and does exist as indicated Min'"

uCi/sec on SPRAD.'

The preferred riheth*d' for decla-,atitn of AGI is',by mimea4ns" otD*Dose-ýsessmhent* using' the MIDAS 'computer' model. ,Howiever, if Kania'hi"onitor reading's "': uist,'ied for longer than 15 minutes and the reqtiifed MIDAS dose absessmentscannoibe confiuplted'within this period, then the declaration can be made using Kaman readings PROVIDED the readings are not from an isolated flow path. If Kaman readings are not valid, field survey results may be utilized.

AGI

Rev. 4

,:,-ABNORMALRAD LEVELS/RADIOLOGICAL EFFLUENT-*CATEGORY" , PAGE 25 of 26 DAEC's- Meteorologic1 Irif6rriati6fihi d Doise Assessmnfit System (MIIAS) was utilized to determine the KAMAIN monitor limits. Eight separate combinations, of release point, source term, mefeorological conditions .ind equipment status were ,analyz.ed. : Pathways considered were the offgas stack, the- turbime build-'g e6lfaiust Vent iund .a single reactor,,

building exhaust vent. Multiple release points were not considered. -In this same vein, it, was assumed that only one of the three reactor building vents is on during the release.

The source terms used have been pre-loaded into MIDAS, and are, the default mixes associated with a loss of coolant accident (LOCA)) and a control rod drop (CRD). The LOCA mix was used in conjunction witha release.via the offgas stack while the CRD'mixi was used for releases via the turbine or reactor building vents: The sou*rce term for a release via the offgas stack is further impacted by the status of the standby gas- treatment ýsystem.

The status of that system was also taken into considerfition.

Based of.1995 data (NG-96-0987), the atmospheric stability was classified as Pascal E'33%

of the time.- Consequentlyjtoth classifications were evaluated.' Based on the same report, the most common wind speeds were: , .

• ' S' ed(I; Pascal Class -Altitde * "Seed n- (mph 156', ,. -;'

'8-12 E , ,1 33' - . . '

Thou~gh ihe temperate setting has no impact on the MIDAS calculations, a value must be entered in order for the program to run. Consequently, the temperature was arbitrarily set at 50 F. ' o .' ' . * . ' ' -,. *. . . V : ' "

The rain estimate was set at zero, to 9liminate any on site washou~of radioactive material.,

For thefsno MIDAS rlinsh a Ciecc concentr'atipn was ass'.ied. -The-resiflts of-these runs were then normahzedio"the limits, -hus' generating atheoreticalKA.MAN limit., Additional MIDAS iuns' weie mrade with ihese theoretical limits ,as ,input,to verfythe nhormalization,

, iputoveif process., ,.

In addition to the total ,integrated dose, MIDAS calculates'a peak whole body DDE-rate resulting from theplume and a peak thyroid CDE rate reslting from inhalation. Because the ASI and AGI KAMAN limifs are to be based on a one hour exposure, establishing, conicentration limits so tfiese peak values match the NUMARC limits is -acceptable.

AGI

- ~ ~ ~ -D04QS~~e~ E9 EBD-A' Rev. 4R.v.4 ABNORMAL RAD6LEVE§S/RADIOLOGICAL EFFLUENT CATEGORY PAGE 26 of 26

...- -' "Site Area- General Initiating Conition. 'Emergency' Emergency

- Cnio . .In ASI" AGI Valid Turbine or RB ventilation'rad monitor,;-,

(KAMAN) reading formore than 15 minutes above: 0.06 gjCi/cc 0.6 g.Ci/cc Valid Offgas Stack "ventilation rad' monitor (KAMAN) reading for more than 15 nihutes above: 40.tci/e. 400 gCi/cc DAEC does not'have a telemetered radiation monitoringsystem., As an alternative, DAEC.

uses valid field survey readings outside the site boundary to determine if doses are greater than 1,000 mR/hr TEDE or greater than 5,000 mR/hr CDE to the Thyroid.

Dose assessment using MIDAS is based on the EPA-400 "methodology, e.g., use of Total Effective Dose .Equivalent (TEDE) and. Committed - Dose Eq'uiv lent - (CDE) Thyroid.

TEDE is somewhat differenf from wholebody' dose from' gas&6is'Sfflu'eit deternii.edlby ODAM methodology which forms the basis for the radiation monitor readings calculated in' AUI. These factors can introduce differences that are at least as large as those introduced by using TEDE versus whole body dose. The gaseous effluent radiation monitors can only detect noble gases. The contribution of iodine's to TEDE and CDE Thyroid could therefore only be determined either by: (1) utlizing the source term mixture in MIDAS, or (2) gaseous effluent sampling. Therefore, DAEC EAL Threýshold Vahie 4 is written in terms of TEDE and CDE Thyroid.

REFERENCES:

1. Offsite Dose Assessment Manual, Section 6.1.2 and 7.1.2, Bases

2. Emergency Plan Implementing Procedure (EPIP) 3.3, Dose Assessment and Protective

, Action , . , - .... '* ,* .. .

3. Radiation Protection Calculation No. 95-001-C, Emergency Actions Levels Based on Effluent Radiation Monitors, January 24, 1995 '- . . ",

4.ý Radiation Engineering Calculation No. 96-007-A, Determination of DAEC Radioactive Release Initiating Conditions for ASI & AGI Emergency Classifications, July 3, 1996,

5. UFSAR Section 11.5, Process and Effluent Radiation Monitoring and Sampling Systems

6. EPA 400-R-92-001,: Manual of Protective Action Guides d'nd Protective Actions for Nuclear Inctidentsk ,., r -r, .:

7. NEI Methodology for Development of Emergency Action7 Levels NUMARC/NESP-007 Revision 4; May 1999.

AGI

EBD-H HAZARDS AND OTHER CONDITIONS -aI AFFECTING PLANT SAFETY CATEGORY Rev.4  !

.-PAGEA.:1f -2b I

.Effective Date:,i: *;,

I JIAIJI .',s, a'

.Date: 0-

• , *.' I a . a 4

"... ,ac.., a ' aa a.'t. I ' ,'a a , '

'a - ' '

• ":*"-- a.

a -

aI a 'I V. .....'a ' , - . a aa aPROCIEDJRE Afr'a'k -OaaV'74 I am responsible for the technical content of this procedure and for obtaining the

-_,r,%!_

I necessary approval from the State and County Emergency Managem'nt officials' prior .

to implementation.

Documentation of State and County Emergency Management approval is via NP 00,"9 a Approved by:. a7Date:a 5;, 7"'

' Manager, Emergency Planning a

'a' - * - r.

-- a-a

'a' a I , ý1

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I I EBD-H:

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORY1, ,PAGE 2 of29

, ,Tableof Contents ..

HUI - Natural and Destructive Phnome'nea Affecting the' Pi6tected Area ................... H-3 HU2 - Fire Within Protected Area Not Extinguished Within 15. Minutes of ... H-6 HU3 - Release of Toxic or Flammable Gases Deemed Detrimental to San of the Plant .............................................. . ,.N..H-7 HU4 - Confirmed. Security Event Which Indicates a Potential Degradation in the Level of:,

Safety of the Plant ........................................ H-8 HU5 - Other Conditions Existing Which in the Judgment'of the EC/OSM Warrant Declaration of an Unusual Event ..............................H... ................ ...................... H-10 b.............

HAI - Natural and Destructive Phenomena Affecting the Plant Vt'l Aaf Area ...- ....... ;. H-1 H"Il -

HA2 - Fire Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown ........................................................................... .H-1 5 HA3 - Release of Toxic or Flammable Gases'ithri a"Facility Structufe Which'Jeopardize" Operation of Systems, Required toMaintain Safe Operations or to Establish or.

Maintain Cold Shutdown ................................................................. H-17 HA4 - Security Event in a Plant Protected Area ..................... "... . ...... L................... H-19 HA5 - Control Room Evacuation Has Been Initiated .......................... H-20 HA6 - Other Conditions Existing Which in the Judgment of the EC/OSM Warrant Declaration' of an A lert ............ ................... ...... ................ .............................................. H-21 HSI- Security Event in a Plant Vital Are ............................ *H-22" HS2 - Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established

........... ..... I ...... .!...............I...................... I...............-.12 HS3 - Other Conditions Existing Which in the Judgment of the ECIOSM Warrant Declaration of a Site Area Emergency . . .............. ... ..................... ............ H-26 HGI - Security Event Resulting in LOSM Of Ability to Reach and Maintain Cold Shutdown

.................................................... I.................................................................... H-27 HG2 - Other Conditions Existing Which in the Judgment of the EC/OSM Warrant Declaration.

of a General Emergency ......................................................................... H-28'

• Rev. 4 HAZARDS ANDOTHER CONDITIONS AFFECTING PLANT SAFETY-CATEGORY' , PAGE 3 6f 29 ,'

HUI Natural and Destructive Phenomena Affecting the Protected Area EVENT TYPE: ,Natural Disasters, Other Hairds and Failures . .. -

OPERATING MODE APPWICA*ILITY: tAl "  :  : 'All EAL THRESHOLD VALUE:

Any one of the following phenomena affecting the Protected Area: ,  ;. ... .t.., , . ... .

1. Valid Amber Design Basis Earthquake (DBE) light and the wailing seismic'alarm on Pan'el1C35,ar&e both activated indicating an acceleration greater than +/- 0.01 gravity. 7

2. Report by plant personnel of tornado striking within protected area boundary.

3. Assessmentby the control room that a destructive event has oc6urred .' ..

4. Vehicle crash into plant structures or systems within protected area boundary that are ddtermined to be Safe ShutdownAreas. . " " a '4'; r- .;. , 14. \ -'

5. Report by plant jgersonnel of an unanticipated explosion within the protected 6'rea-ldundary resulting in visible damage to perman6iitksiructures or equipment requied for Safe Shutdown,

6. Report of turbihe" , failure

"- resulting a .. . . ina-*.

. casing penetiation a " 6r damageto

- a5 turbine or- generator ,' - _ -' seals. " , ,

7. River flood water levels above 757.0 ft. ,

a .

8. The Max Normal operating water level exceeding and EOP 3 limiis.,

• t" . J

9. River water level below 725 ft. 6 in. ,, , ... - : ... .;

DAEC EAL INFOR_,ATION:. a... ,,, . , ", ,,.

EAL Threshold Value I addresses earthquakes that are detected in accordance with AOP. 901. For DAEC, a minimum detectable earthquake that -is, indicated on panel, 1C35" is an acceleration ,,greater.than .- ,0.01 Gravity.,

DAEC EAL Threshold Value 2 addresses report of a tornado striking within the protected area or within the plant switchyard. ,. ":.. -,.

DAEC EAL Threshold Value 3 allows for-the control room to determine that and event has occurred and take appropriate action based on personal assessment as opposed to verification. No attempt is made to assess the actual magnitude ofth"edanihge. Such damage can be due to collision, tomadoes,'missiles,,or any other cause. Damage can be indicated by- report to the control room, physical ,observation,- or by Control Room/local control station instrumentation. Such items as scorching, -cracks, dents, :or discoloration of equipment or structures required for safe shutdown are addressed by'this EAL.

HUI

DAEC EAL Threshold Valiie 4-Addr'esses'a vehicle (automobil6, aiicrAfl, forklift truck or ftain') crasl hat may potentially damage plant,structures, containing functions'and'systems required for safe shutdown of the plant. This does not include. Vehicle crashes with erch other'or d-arnage t6o office? or -ar'eliouse- structuris Escalation, to Alert under HAI would, occur if damage was sufficient to' affect the ability to achieve or maintain safe- shutdown, e.g.,, damage, made ,required equipment inoperable' '*osfriichir*a*driage 5 w-a observed such as bent supports' or pressure boundary leakage. '

Safe Shutdown Areas "Category.. ... "-.'Area Electrical Switchyard, 1G31, DG and Day Tank Rooms; 1G21!DG did Day Tank Power Rooms, Battery Roonms; Essential Switchgear Rooms, Cable Spreading Room Heat Sink/ Torus Room, Intake Structure, Pumphouse C oolant ... . .... ... . ': .

Supply . .. .. . ..-- . ....... ,

Containment Drywell , T s, , .. . ..-.-- ,

Emergency NE, NW, SE Comer Rooms, HPCI Ro6rm RCIC Room; RI-R Valve, Systems Room, North CRD Area, Sotith CRDIArA , ,

Other Control Building, Re6inot& Shutdown Pa 'elIC388 Area Aiea,' SBGT Roomn m - .Pan-ll-.6 - a DAEC EAL Threshold Vaiue5 adrdrsseixllosions wiihin the protected area. As used here, an explosion is a rapid, violent, unconfined c6fibusfion, or a catastrophic failure of pressurized equipment, that' potentially imparts significant energy to, near-by, structures or, equipment. Damage can be indicated by report to the cb6fitrol' roo*mph~i6a'bbseaion, or by Control6'Room/local contiol station instruinentation..

Such 'items as scorching; cracks; dents,, "o 'digcoloration of equiipmenti or structures required for safe' shutdown are addressed by this EAL. The EC/OSM needs to consider the security aspects of the explosion, if applicable.

DAEC EAL Threshold Value 6 addresses turbine failure, causing observable damage to the turbine casing or damage to turbine or generator seals..... , to t DAEC EAL Threshold Value 7 address~es the observed effects of floodingin.accordance.with A)P 902.

Plant site finished grade is elevation 7'57.0 ft.' Personnel doors and railroad and truck openings at or near grade w6uld"require prdtecti6n'in ilt*e ve-nt'ofa flood above elevation 757.0 ft. Therefore, EAL 7 uses a threshold of'flood water Ievels'above 757.0 ft.

DAEC EAL Threshold Value 8 addresses internal flooding can be due to system malfunctions, component failures, or repair activity mishaps (such as failed freeze seal) that can threaten safe operation of the plant.

HUI

A, Rev. 4 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORY- PAGE5 61f29 -'

Therefore, this EAL is based on a valid indication that thewater level is'higherthan the maximum iohiral, operating limits. The, Maximum Normal Operating Limits are defined as the highest values of the identified parameter expected to occur duringnormal plant operating conditions with all directly associated support:

and control systems functioning properly. Exceeding these limits ,is an lentry- conditioh' into EOP 3, Secondary Containment Control and may be an ipdication thatywater~from. a primary system is discharging, into secondary containment. Exceeding the maximum normal.operating limit is interpreted as a potential degradation in the level of the safety of the plant and is appropriately treated as an Unusual'Event, emergency classification. The maximum normal operating water level limits are taken from AOP 902 and EOP 3 and are shown in the table below:... ... , . ', **-... -. - '--

.. _.. ', - " -. r'. -*,.

,.? . .'--y. t _- ,-,r -. ., ...

Maximum Operating Limits - Water Levels Affected Location Indicator Maximum, Maximum Safe

_Normal OL OL HPCI Room Area - '113768- - 2 inches 6 inches RCIC Room Area LI 3769 '3 inches 6 inches A RHR Comer Room SE Area LI 3770 2 inches 10 inches B RHR Comer Room NW Area  : '1 LI-3771 - 2 inches -- - 10 inches '*'.

Torus Area -' . ', ..... LI 3772 2.inches , ',, - 12 inches

ýEAL Threshold Value 9 addiesses the effects of low river water level.' The intake structure for the safety-'

related water supply' systenis (river water, R R:sbc RHR-ervce wter wat'-"andandemegeny

.. .rg iieservicew

" ,wýe)te)isis locate-d atedoon the west baink of the'Cedar River. -Aii6verfl -typd barrier across the fiver was designed -ahd constru6ted in accordance With SeismicCategory I criteria 4o-intercept'the stream'b&d flow n?6id diierf'it to the intake' structure. This makes'the'entire'flow of the river available to the safety-related'wat'ersiipply' sy-stems. ý'A minimum flow of'13 'cubic feet per second (cfs) from a minimum I000-year river floiW of 60 6fs must be diverted. The top of the barrier wall is at elevation 725 ft. 6 in.' :River water'level below this level iejpresehts a potential degradatin m the level of safety'ofthe pl and is addressedby EAL Threshold Vaue9.

In this EAL, "Vital Area'sis-defined'as plant stru6tures or areas contammigiequlpment necessary for a ssfe shutdown. ', "' "' . .. ... )...

REFERENCEFS: - " * :*,.-. ,. ,*, ,.,,.., , ,: " _ , . , . o 1..-Abnormal Operating Procedure (AOP) 901; Earthquake' - ' ' '

2. Abnormal Operating Procedure (AOP) , Flood90 ' . " "

3. Abnormal Operating Proc~diwre'(AOP) 903,'T7rnfadb "'" ' "

4. Emergency Operating Procedure (EOP)-3, Secondary Containment Cofitrol ' ' '

5. EOP Basis Document, EOP-3, Secondary Containment Control6 i':

6.. UFSAR Chapter 3, Design of Situctures, Coinponeents, Equipiiiei, ad Systemis -

7. Bechtel Drawing BECH-M017, Equipment Location - Intake Structure Planis at Elevations, Rev. 6 HUI

HAZARDS AND OTHER CONDITIONS'AFFECTING PLANT SAFETY CATEGORY,, 'PAGE 6, of 29' HU2 Fire Within Protected Area Not Extinguished Within 15 Minutes of Detection EVENT TYPE: Fire "

OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

1. Fire in buildings or areas contiguous to any of the following areas not extinguished within 15 minutesof' control room notification or verification of a control room alarm:

• Reactor, turbine, control, admin/security S-Intake structure "' " °.

• Pump house DAEC EAL INFýORMATION:.- " 9,, - J The purpose of this EAL isý t6 addfe*s the -magnitude and extent of fires that may be potentially significant precursorsto damage to'safety systdms. This includes such iteims as fiies within the'administration building,,

and security building (buildings contiguous to the reactor building, turbine building and control building),

yet, excludes fires in the warehouse or construction support center, waste-basket fires; and other small fires of no safety consequence. As used, here,:.Detection is visual observation and report by plant personnel or:

sensor alum, i1dication* The 15 minute, time. period begins with a credible -notification. that. a FIREis.

occurnng,,or notification of a-VAL. D ýire,,etec.on system alarm, Y.eification, of a,fire detection system alarm includes actions that can be taken within the control room or other nearby locationto ensure that-the' alarm is not spurious. A verified alarm is assumed to be an indication of a FIREunless it is disproved within the 15-minute period bypersonnel dispatched to the scene, In other words, a personnel report from the scene may be used to disprove a sensor alarni if received within 15 minutes of the alarm. -9

, , , , -,,. , * , .- , , .. ", . .- *. {: . .

Per AOP 913, the location of a fire can be determined by observingl,40B.alarm messages, Zone Indicating Unit (ZIU) alarms, or fire annunciators on panels 1C40 and IC40A. 1, *he location of a fire' can' also be determined by verbal report of the person discovering the fire. Verification of the alarm in this, context means those actions taken to determine that the control room alarm is not spurious.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 913, Fire .- -

2. Abnormal Operating Procedure (AOP) 914, Security HU2

I' EBD-H Jl ,

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORY

.Rev. 4 .

PAGE 7 of 29

. i HU3 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant EVENT TYPE: Other Hazards and Failures I A OPERATING MODE APPLICABILITY: All I-4 EAL THRESHOLD VALUE: A.. - *.

Safe operation of the plant is jeopardized by one of the following:

'1. Report or detection of toxic or flammable gases that could enter within the site area boundary in amounts that can affect normal operation of the plant. .

2. Report by Local, County or State Officials for tiotential evacuation 6f site& eisbiinel bas'ed 6i offsiteV event.

,,.DAEC EAL INFORMATION: - -  : '

This -Threshold Value is based on releases*in 'coricentrations -fliin thd site lboundarr that will affect the health of plant personiel' or -affecting the safe loPeratio -ofthe plant' with the planft being Vwithin -the eVacuati6n area of an offsite event (i.e., tanker trick ccid etc) The evacuation area is as determined from the'DOT Evacuation Tables for Selkeed Hazýýd aterialin the DOT Errirgency

,Response Guide for Hazardous Materials.. .; .., L.

For the purposes of this ,EAL; CO2 (such as~is discharged by the fire tkspr n system) is nrot toxic. CO2 can be lethal if~it reduces :oxygen-to 16W conceititatibfis that are ifiniediateIi d*llg&r6iis to 'life'and'health (IDLH). C0 2 discharge into an area is not basisfor emergency classification under this IC unless: (1)

'Access :to the affected area is re~juired, ;aid-(2) C concentrdtion results refconditions thai mai'kthe area uninhabitableorinaccessible(i. L .- )": " , *. ." , "; , " ..

REFERENCES:

.. ,. -,: ' I

1. UFSAR Section 2.2, Nearby Industrial, Transportation, and Military Facilities

2. UFSAR Section 6.4;,Habitability Systems - -- -A " - '. S *,.

V *A V A HU3

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORY,, PAGE 8 of 29".

HU4 Confirmed Security Event Which Indicates a Potential Degradation in the Level of Safety of the PlAnt EVENT TYPE: Security event with potehtial loss of level of safety of the plant.

OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE: '

One of the following:

1. Suspected sabotage device discovered within the Protected Area AND outside a plant Vital Area.

2. Suspected sabotage device discovered outside the Protected Aieaiih the plant switdlyaid.'

3.. Confirmed tanpering*

6ý th sfet' r -a1' equipmfen. -- !-'! - .-. " .

4. A hostage situation that disrupts nbrinal plant opeiations.

5. Civil disturbance OR strike which'disrupts normal p operiations.

6aiit

6. Internal disturbance that is not short lived or that is not a harmless outburst involving one or more individuals within the Protete~d Area., '

7. "LW" credible threats as determined by NMC SE-0018, "Security Threat Assessment".

DAECEAL INFORMATION:"

Security events which do not represent at least a potential degradation in the Ie1e'l'of safety of theplant aie reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72., The term "suspected sabotage device" is used in place of "bomb device" for consistency with the DAEC Safeguards Contingency Plan.

Consultation with Security supervision is required to determine these Threshold Values.

EAL I describes a suspected sabotage device discovered within the ProtectedArea but outside an area that contains safety functions or systems. It is a potential degradation of the level of safety of the plant and is an UNUSUAL EVENT.

EAL 2 describes a suspected sabotage device discovered in the plant switchyard representing a potential degradation of the level of safety of the plant.

HU4

I EAL 3 is for confirmed tampering and is adapted from the list of security, plan contingencies. I EAL 4 identifies a hostage situation that disrupts normal plant operations. A hostage situation is -

considered to disrupt normal operations if it results in the inability to perform surveillance activities, alters unit operations, or as described in the security plan.

EAL 5 describes a civil disturbance or strike is considered to be a spontaneousactivitythat disrupts normal plant operations. A civil disturbance or strike is considered to disrupt normal plant operations if it initially disrupts normal ingress or egress to the owner controlled or protected area, or if it requires_

assistance from the Local Law Enforcement Agencies (LLEA) to control. .

EAL 6 deals with suspicious internal disturbances that may have been planned by unauthorized personnel as a diversion to gain entry to the site property. ..

EAL 7 ensures that appropriate notifications for te., security. theatare madeein.a -timely manner. The determination of"LO credible threat" or "HI credible throat" is based on information found in NMC SE 0018, "Security Threat Assessment". The emergency response to a "LO" Credible Threat is initiated, "through AOP 914, "Security EN;ents" and EPIP 2.8, "Security 'Threat'S: 'A'"HI credible" threat would escalate this classificatioii to the ALERT status as an J1A4. Only the plant to which the specific threat is made need declare the Notification ofUnusual Event.  :.

Suspected sabotage devices discovered within the plant Vital Area would result M" escalation via other Security EALs.

S

REFERENCES:

. °* -, ,.:,

1. Abnormal Operating Procedure (AOP) 914, Security Events '. " .,

2. NMC SE-0018, "Security Threat Assessment"

3. EPIP,2.8, "Securityjhreat" ., .: . . -' - r.. -r . .

S! '3:.

,I HU4

'EB D-H . ,*

Rev. 4 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORYi" fýPAGE, 10 of 29. \-'

11U5 Other Conditions Existing Which in the Judgment, of the EC/OSM Warrant Declaration of an Unusual Event EVENT TYPE: EC/OSM Judgment'.

OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

1. Other conditions exist which in the judgment of the Emergency Director indicate a potential degradation of the leyel of safety of the plant.

9 9' DAEC EAL INFORMATION:

The EAL addresses conditions that fall under the Notificatiodn'of Unusual Event-emergency classification descripti5 contained in NUREG-0654; Appendix 1, that is retained under'the generic methodolog.

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 degradatio6nof safety systenm"s 'ccurs.'

Per EPIP 2.5, the Emergency Coordinator/Operations Shift Manager (EC/OSM), is, the title for'the emergency director function at DAEC.

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation

2. NUREG-0654/FEMA-REP-l, Criteria for Prepariton and, Evaluation of 'Radiolbgical,Emergency Response Plans'and Preparednesi'in'Suiport of Nuclea&ý "Pbier' Plaiits,'Revision 1, October. 1980, Appendixl , ,

HU5 HU5

L~8A~1~UMNP" BD-H Rev. 4 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORY-' 'PAGE 11 of29' hA1 Natural and'Destructive Pheno ena Affeding the Plant Vital, SArea . , .- * ""

EVENT TYPE: Natural Disasters, Other Hazards and Failures .

OPERATING MODE APPLICABILITY: All' '

EAL THRESHOLD VALUE: ,

Anyone of the followig phdnom~na affecting the Protected Area: ;i ""

1. !Valid Amber .Operating Basis Earthquake (OBE) light and the wailingý seismic alarm on Panel 1C35 are

- both activated indicating an acceleration greater than _ 0.06 gravity - ' - " "

2. Tornado striking plant vital areas.

3. Assessment by the control room that damage has affected Safe Shutdown Areas.

4. Vehicle'crash affecting plant tvitl aireas.""

5. Sustained high wind sjpbed of 95 'riiles per hoiir or above'affecting'plant vital areas.

6 'Missiles affecting safe shutdown areas. ' , .

7. Rive'r dflobd Water lev6ls'above 767.0 ft. ..' .'

8. The Mai'Szfe operating'water level exceeing and EO 3limits in two or, more are shutdown is required- - . . ..... a

9. River water level below 724,ft. 6in.-,K -,, ' ", - . ... , . . '

• " , I -

DAEC EAL INFORMATION: ""

There are no significant deviations from thegeneric EALs. For the events of concern here,"'the key issue is,..

not the, wind speed, earthquake:intensity,-'etc.;f but' ivhether -there'is ,reultahtdaWnag-e"to equipment 'or structuresrequiredto achieve or maintain,safe shutdown, regardless offthe cause,Determination of damage affecting the aility to achieve or maitaimn safe ,shutdown can be indicatedbyreports to the control room,e physical 6bseivation'or y"Contr6ol Room/local control station instrumentationi. -

EAL Threshold Valde I addresses OBE events- that are detected ir566ordafice with AOP 901. For DAEC,'-"

the OBE is associated with a peak horizontal acceleration of+ 0.06 Gravity.  ;. . . '

DAEC EAL Threshold Value 2 5addre'ss"ssr~i6it'6f atonmado striking a plant vital area.

HAI

BSSD0OMNEBD-H ,

S ~~~Rev'.4, Li I HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORY PAGE 12 of 29" DAEC EAL Threshold Value- 3, addresses a report to, the control room of damage affecting safe shutdown areas. The reported damage can ble from thmadoes, high winds,'fooding;, missiles, collisions, or any' other cause. . , ,!

DAEC EAL Threshold Value 4-addresses vehicle (automobile, aircraft, f6rkdift,, truck 0i" train) confirm-ed crashes affecting plant vital areas. This does not include vehicle crashes with each other or damage tb office or warehouse structures. , ,

DAEC EAL Threshold Value 5 addresses sustained high wind speeds as mefisured by the 33-Foot or 156 Foot elevations on the Meteorological Tower. Sustained wind.speed means the baseline wind speed measured by meteorological tower that,does not, include gusts.' The design basis wind speed is 105 miles per hour. However,,, the meteorological instrumehtation is only ,capable 6f measuring wind, speeds up to 100 miles per hour. Thus the alert level for sustained high wind, speed, 95 miles per hour, is selected to be on scale forthe meteorological instrumentation and to conservatively account for potential mieasurement errors.

DAEC EAL Threshold Value 6 addresses missiles affecting safe shutdown areas.ý Such missiles can be from any. cause, e~g., tcmado-generated; turbine,' pump or other rotating machinery catastrophic failure;. or generated from. an explosion. ,

PerAOPs 913 and 914, the following areasare identified as,safe shutdown areas and are shown on-the EAL tables. This table is displayed as an aid to the Emergency Coordinatorin determining appropriateareas of concern. , - , -

Safe Shutdown Areas Category , , - Area Electrical Switchyard,, I G31 DG and Day Tank Rooms, 1G21 DG and Day Power Tank Rooms, . -: ,

Battery Rooms, Es§-ential Switchgear Rboims,, Cable Spreading Room ,; . , . .. .. . . - . _ .

Heat Sink/ Torus Room, Intake Structure,, Pumphouse, ,:,

Coolant S u p p ly ... . ...

Containment Drywell, Torus ....

I -

r-mergency NL-, NVV, EE Corner, ooms, .P I Room, RClC Room, RHR Systems, Valve Room., North CRD Area. South CRD Area"',

Other Control Building, Remote Shutdown Panel 1C388 Area, Panel .4I 11C56 Area,,SBGT Room ,,.. - ,,, ,- j I DAEC EAL Threshold Value 7 addresses river water levels exceeding design flood water levels. All Seismic Category I structures and non-seismic structures housing Seismic Category I equipment are designed to withstand the hydraulic head resulting from the "maximum probable flood" to which the site Ky HAI

ppnI-_

(

HAZARDS'AND OTHER CONDITIONS AFFECTING PLANT SAFETY'CATEGOR-Y 'PAGE 13'of 29" could be subjected. The desigil f-ood water is at elevation'767.0 ft: 'Ma'jor equipriment 'p'*iietrati6ns in ihe exterior walls are loiated above elevation 767.0 ft. Openings'belotv the flood level are eitfier watertight or*

are provided with means to control the inflow of water in order to ensure that a safe shutdown can~b6 achieved and maintained. Consideration has also been given to providing temporary protection for openings in the exterior walls up to flood levels of 769.0 ft. All buildings were also'checked for uplift (buoyancy) for a flood level at elevation 767.0 fl, and the minimum flictor of safety used was 1.2.' Theref're, DAEC EAL 7' uses as its threshold flood water levels above 767 feet.

DAEC EAL Threshold Value 8 addresses internal 'flooding consitent with the" r"cjuirements of BEOP 3, Secondary Containment Control. If RPV 'pre'ssure-'reduction -'ill *decrease*6lea'age' into secoridary containment then this is due to leakage from the primary-systebm,?which i's addressed by the Fission Barrier Table indicators and System MalfunctionEALs, and is not addressed here. Therefoie, EAL 8 addresses' conditions in which water level ini two or more areas is abobe Maxinium Safe'Operating Limits and reactor shutdown is required.'Requiredmeans that the reactor shutdown wasI procedurally mandated by EOP 3 and is not merely performed as a precaution or inadvertently. Maximum Safe OperatingLimits are defined as the highest parameter value at which neither (1) equipment'necessary',for-safe shutdown of the plant will fail nor (2) personnel access necessary for the safe shutdown of the ilant'will.5e~pi'66liidd'. 11.The-intemal flooding can be due to system malfunctions, component failures, or repair activity mishaps (such'as-failed freeze seal) that can threaten safe operation of the plant. This includes water intrusion on equipment that is "notdesigned to be submerged (e.g., motor control, centers)..';"

The maximum safe operatingwater level limits are takenfrom EOP 3 and are shown on the table below:

Maximum Operating Limits - Water Levels,', '

Affected Location Indicator -'Maximum Maximum Safe

- c... ",.NornialOL - .:j .

HPCI Room Area LI 3768 2 inches 6 inches RCIC Room Area , . ".LI'3769-. , 3 iniches, . 6 inches

-l A RHR Comer Room SE Area LI 3770 . . 2_inches - 10 inches, B RHR Comer Room NW Area L13771- '"2inches ,10inches Torus Area LI 3772 2 inches 12 inches DAEC EAL Threshold Value 9 addresses the effects, of lowjriver.water level. The intake structure for the safety-related water. supply 'sstes' '(ri'ii wvdter,'RHR servic&-iwater, -anid emergency service water) is located on the west bank of the Cedar River. Thbov6ifloow Weir is it elevatin 724 feet 6 inches. River level at or below this elevati6h.*will resultih aill fiver fl]6xa eing diverted to the'safefy related water supply systems. The top of the intake structure around the pump wells is at'&levation' 724 feet. If the river water level dropped to this level, the pump suction would have no continuous supply. Therefore, this EAL uses a threshold of water level below 724 feet 6 inches as a potential substantial degradation of the ultimate heat sink capability. I HA1

HAZARDS AND OTH-ER CONDITIONS AFFECTING PLANT SAFETY CATEGORY, In this EAL, "Vital Area" is defined as plant structures or areas containing a safe shutdown.

• 1 k*, - . 1*,.o , * * ° ", , . *- ' * *

1. Abnormal Operating Procedure (AOP) 901, Earthquake

2. Abnormal Operating Pr66edure (AOP) 902, Flood

3. Abnormal Operating Procedure (AOP) 903, Tornado

4. Abnormal Operating Procedure (AOP) 913, Fire

5. Abnorniial Operating Procedure (AOP) 914, Security Events

6. UFSAR Chaiter 3, Design of Structures, Components, Equipment, and Systems

7. Bechtel Drawing BECH7 M01l7, Equipment Lochtioin - Intake Stiucture Plans at Elevations, Rev. 6

8. EOP Basis Documefit, EOP 3 - Secondary Containment 'Control Enimrgency Operating Procedure (EOP) 3, Secondary Containment Control

").

j S S -.

i HAI

AL"BASES. DOA EBD-H HRev.4 HAZARDS AND OTHER CONDITIONS AFFEC-TING PLANT SAFETY CATEGORY PAGEl15 of 2-96 HA2 Fire or Explosion Affecting the Operability of Plant Safety Systems Required to Establish or Maintain Safe Shutdown EVENT TYPE: Fire

• 4*

OPERATING MODE APPLICABILITY: All I *, 444 EAL THRESHOLD VALUE: - -' ' ,". ' ,:

1I Fire ori explosion affecting one of the following systems or areas of concern. -

, 'SYSTEMS 1/4 - , . -

44 4 . . .4 -t

• Reactivity Control S-o Containment (Drywell/Torus)..

• HRHRCore Spray/SRV's I -

S-, PCIRCIC

• RIRSW/River Water/ESW

• Onsite AC Power/EDG's Offsite AC Power . '

• InstrumentAC .

• -DC Power * ' * ' 4: 4 .44 Remote Shutdown Capability--..

44:

AREAS

• Reactor, Turbine; Control, Admn/Security

", Intake Structure " " -4' '4

'4- -. 1

+' t+ +

Pump' Hbuse ..- ' l +" " . J 4-,AND,

2. Affected system parameter indications show degraded performance or plant personnel report VISIBLE..

DAMAGE to permanent structures or equipment within the specified area.

DAEC EAL INFORMATION: 14 1.. -

There is no significant deviation from the generic EAL. Of particular concern for this EAL-'are fires that may be detected in the reactor building, control building, turbine building, pumphouse, and intake structure as shown in Tabs I and 3 of AOP 913. Damage from fire or explosion can be indicated by physical 4 HA2

HAZARDS AND OTHER CONDITIONS AFFECTING PLANTSAFETY CATEGORY - "PAGE 16 "f 2 observation, or by Control Room/local control station instrumentation. No attempt is made in this EAL to assess the actualmagnitude ofh d Per AOP 913, the Iocation of a fiie canb'e determined by obserying IC40B alarm messages, Zone Indicating Unit (ZIU) alarms, or fire annunciators on panels IC40 and IC40A.

2 I NOTE:

Scope of Systems and Equipment of concern established by review of Appendix R Safe Shutdown credited systems. Only those systems directly affecting safe shutdown or heat removal are listed for consideration, due to fire damage. Support System's a-d equipment such" as' HVAC and sj ecific instrumentation, while included in Appendix R analysis is not considered an immediate threat to the ability to shutdown the plant and remove decay heat.

This EAL addresses a FIRE / EXPLOSION and not the degradation in peformance System degradation is addressed in the System Malfunction EALs. The refeience to damage" of affected systems.

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 operationI Th'&significaiice here is notthat a safety system was degraded but the fact that the FIRE / EXPLOSION was large e"'ough to cause damage to these systems. Thus, the designation of a single train was intentional and is appropriate when the FIRE /

EXPLOSION is large enough to affect more than one component. Lagging fires, 'fires in waste containers or any miscellaneous fires that may be in the vicinity of safety systems, but do not cause damage to these systems, should NOT be consid(Fied f6i-tliisEAL. .d a th With regard to EXPLOSIONS, only those EXPLOSIONS of sufficientforce to damagepermanent structures or identtied equipment requiredforsafe operation,shou' be considered. As used here,, an EXPLOSION is a rapid, vmiolent;,! unconfined cbmbustnio, or a catastrophic fufiiur of pressunized equipment, that potentially imparts significant energy .to iiear-bystruciuies 'and' materials. ,',e occ~urrenceof the EXPLOSION with reports of evidence of damf'age (e.g.,)deformation, scorching) 'isufticient"for the,,declaration. The EC/OSM also needs to consider any security aspects of the EXPLbSIONS, if apiliba'ble'"

REFERENCES:

1. Abnormal Operating Procedure (AOP) 913, Fire

2. Abnormal Operating Procedure (AOP) 914, Security Events "

3. Abnormal Operating Procedure (AOP) 915, Shfitdo"Vxi Outside'Contr61 Room

4. UFSAR Section 6.4, Habitability Systems HA2

S . L/\\:E-

. ... I .... ... EBD ev HAZARDS AND OTHER CONDITIONS AFFECTING PLANTSýAFET'Y-C-AT-EG'ORY Rev. 4 PAE 17 of 29 I

m 11A3 Release of Toxic or Flammable Gases Within a 'Faidli'ty Structure Which Jeopardizes Operation of Systems Required to Maintain Safe Operations or to

-Establish or Maiintaifi Cold Shuidow n EVENT TYPE: Other Hazards and Failures OPERATING MODE APPLICABILITY: All -*..

EAL THRESHOLD VALUE:,.

One of thefollow g

1. Report or detection of toxic, gases within a Safe Shutdown Aiea 6 entrations

'con that will be life threatening to plantto personnel.

n rsone" f""',_ .

• _

OR - "'

,_2.Report or detection of fldm'niabl gases*wii a "f' d *

Raeportordection oft Shutdown Area in concentrations that will affect the evacuatioprin accordance with AOP 915. - ... .. I I I Foepurposso EhisL@,e CO2'(sbch~as. is.'d-&chargid by the fire su'ppression system) is,not toxic. C02

_,rcte "ocnrto I '"

can be lethal if it reduces oxygen to 10w* con centrations that are lmmnediately, dang'erous -to life and health (IDLt.-1. C02 dischag'e 'into'a'n'are~ais not "basis'for emergbnz~lýýassificatt'on under this IC unless. (1) ccess ,o thi a to area is requAred, ansd (2)ad Cr oentr ftoi resulsin cotnditions that make the area ecnhabtitableor dancessble w (i.h AP 9DLH1).

TOXIC - Exposure to the worker in excess of the limits specified in 29 CFR 1910.1000. In practice, this should be considered for concentrations which are capable of producing incapacitation of the worker.

cne source of the release is NOT of imm e d iate concern forthase threshold values. The concern is for the health and safety of plarntotain aPleant thea in a safe operating condition.

This EAL is based on gases that have entered plant structures that will affect the safe operation of the plant. These structures include buildings and areas contiguous to plant vital areas and other significant HA3

-EBD-H Rev. 4 .1 HAZARDS" AND OTHER 1 .. 1. CONDITIONS

' ; I -j.. , . AF.FECTING PLANT SAFETY C.TEGORY- pAGE ._ G 18P 8 f229 K)

I buildings or areas. The intent of this EAL is NOT to include buildings or 6ther areas that are NOT contiguous or immediately adjacent to plant vital areas. -"

Per AOPs 913 and 914, the following areas are identified as safe shutdown areas. This table is displayed as an aid to the Emergency Coordinatorin determiningappiopriateareasof concern. I Safe Shutdown Areas Category Area Electrical Power Switchyard, 1G31 DG and Day Tank Rooms, IG21 DG and Day Tank Rooms, Battery Rooms, Essential Switchgear Rooms, Cable Spreading Room,s ,-. *I Heat Sink/Coolant Supply Torus Room, Intake Structure, Pumphouse Containment Drywell, Torss Emergency Systems NE, NW, SE Comer Rooms, HPCI Room, RCIC Room, RHR'

" Valve Room, North CRD Area, South CRD Area Other Control Building, Remote Shutdown .Panel -1C388 Area, Panel I C56 Area, SBGT Room , , -,;

REFERENCES:

1. Abnormal Operating Procedure (AOP) 913, Fire

2. Abnormal Operating Procedure (AOP) 914, Security Events

3. Abnormal Operating Procedure (AOP) 91 5, Shutdowr. Qutside Control Room-,

4. UFSAR Section 6.4, Habiiability,

+, ) 4"* , * * ~. I*ýSyems -,

'h$ .- ' -*' ii -.

r -

1 -- 'I,

- I- 'f I;.-

rr .1 -

IN I- I I K

HA3

EBD-H Rev. 4 HAZA;RDS AND OTHER CONDITIONS AFFECTING PLANT SAFETlYCATEGORY PAGE 19 of 29 I

1A4 Security Eventin a Plant Proteed Area"-' :--.... ': --

EVENT TYPE: Security

- 7 - -7 - - *7A OPERATING MODE APPLICABILITY: All 777 j7 - 7 7 -, - -

EAL THRESHOLD VALUE

• Pt # &

One of the following: - .... ,."

1. Intrusion into plant Protected Area by a hostile force. , ,

1 ý I

2. Any security .vent of increasing severity that'pe'ristsfor30 minutes:

a. Credibklebomb threats '- -

b. xtorton.

7 .: - -

c. Suspicious Fireor Explosion', :. .ir,,-,: * ... 1.7

d. 'Significanit Security System Hardw~areFaifuir' ' J?

e. Loss of Guard Post Cdntact "

"3. "HI" Credible Threats as determined byNMC SE-0018, "Security Threat Assessment"...-?.

I DAEC EAL INFORMATION: 7 40 EAL I is an intrusion of a hostile force into the Prbtected Areaý repr's67tifig ýXote "

rersnu gap ubstantial degradation of the level of safety of the plant. A civil disturbance, which penetrat & cted Area, can be considered a hostile force.

EAL 2 security events represent an escalated threat to plant safety above that contained in the Unusual Event Under this EAL, adversaries within the protected area are not yet affecting nuclear safety systems, engineered safety features, or reactor shutdown capability that are located within the vital area. A security event is considered to be "of increasing severity" if events are NOT under control of the security force within 30 minutes. Intrusion into a vital area by a hostile force will escalate this event to a Site Area Emergency.

EAL 3 is the determination of "HI Credible Threat" based on information found in NMC SE-0018, "Security Threat Assessment". The emergency response to a "HI" Credible Threat is initiated through AOP 914, "Security Events" and EPIP 2.8, "Security Threat".

REFERENCES:

1. NMC SE-0018, "Security Threat Assessment"

2. Abnormal Operating Procedure (AOP) 914, Security Events HA4

'*i EBD-H HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORY.PAGE 20 of 29 HA5 Control Room Evaquation Has Been Initiated, EVENT TYPE: Control Room Evacuation OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

1. Entry into AOP 915 and initiation of control room evacuation.

DAEC EALINFORMATION:, - -'

The applicable procedure for control room evacuation at DAEC is AOP 915.

Evacuation of the Control Room represents a potential for substantial deqridation of the level of safety of the plant. and therefore requires dn' ALERT declarat1'?.6 ýAdi tional support, monitoring and direction is required and accomhlished by ciivati6on of the-Techmnical SupportCenter at theALERT classification level.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room

2. UFSAR Section 6.4, Habitability Systems HA5

1 1 EBDI.-H

- ~Rev. 4 LHAZARIDS AND OTHER CONDITONS AFFECTING PLAN'T -.S'AFETiY

'C"AT"'E~gGORY PFAGE21Iof29.

HA6 Other Conditions Existing Which in he'Judgment of ihe EC/OSM Warrant Declaration of an Alert EVENT TYPE: EC/OSM Judgment  :

OPERATING MODE APPLICABILITY: All ,. ., .

EAL THRESHOLD VALUE:

1. Other conditions exist which in the Judgment-of the Emergency Director/indicate that plant-safety7 systems may be degraded and that increased monitoring of plant functions is warranted.

DAEC EAL INFORMATION::

Events are in procesis or have octiiffaewidti iv o.ve,,0, actual orpoteftial, substantial degradatibn of the level of safet of the h .EPAProtectivept Action Guideline expostirelevels. " " ",d' s fra"io of t Arti

_.Per EPIP 2.5, the Emergency Coordinator/Operations Shift Manager (EC/OSM) is the title for the emergency director function at DAEC. The EAL addresses conditions that fall under the Alert emergency classification

]

description contained in NUREG-0654, Appendix 1. I, , ,

REFERENCES:

1. Emergency hPln Im enting Procedure (E 2.5, Control Room Emergency Response Operations

2. NUREG-0654/FEMA-REP-1, Criteriafor Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of.Nuclear Power Plants, Revision I,- October 1980, Appendix 1 HA6

HAZARDS AND OTHERCONDITIONS AFFECTING PLANT SAFETY CATEGORY-, PAGE 22 of 29 HSI Security Event in a Plant,Vital Area EVENT TYPE: Security OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

One of the following:

1. Intrusion into plant Vital Area by a hostile force.

2. A security event which results in the loss of control of any Vital Area (other than the Control Room).

3. IMMINENT loss of physical control of the facility (remote shutdown capability) due to a security event.

4. A confirmed sabotage device discovered in a vital area.

DAEC EAL INFORMATION:

IMMINENT - Mitigation actions have been ineffective and trended information indicates that the event or condition will occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

This threshold value escalates from the ALERT Protected Area intrusion to a Vital Area intrusion of a hostile force.

A security event is as defined in the Safeguards Contingency Plan.

Loss of physical control of the Control Room OR loss of physical control of the remote shutdown capability due to a security event, is to be classified as a GENERAL EMERGENCY per Initiating Condition HGI.

A "confirmed sabotage device" is a determination made by the security force through the Security Plan, Contingency procedures and other guidance documentation.

This class of security events represents an escalated threat to plant safety above that contained in HA4, Security Event in a Plant Protected Area, in that a hostile force has progressed from the Protected Area to the Vital Area. Under the condition of concern here, the adversaries are considered to be in a position to directly and negatively affect nuclear safety systems, engineered safety features, or reactor shutdown capability.

HSI

EBD-H Rev. 4 296 FHAýZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CAT'EGOR6Y' PAGE 23'6f

REFERENCES:

'- - ., -'- - F F

F 4, .Fk 1/2 -

F - F -. -I.

1. Abnormal Operating Procedure (AOP) 914, Security Events

2. NEI Methodologyfor Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May.

1999 ' - ,

'F - F F F - F

• F 1

• . -- t * £. OL F' .* I ) 'j C -- r  :' - -'. "F

'I - - F -.

'F *' 'F'

- . - 'F.' -

¶ .

lit. 'Fn. *,dF F IF .',., F'F.F' F:

F,' ,

- $ F -. j'. * '!'

F ' .

- Fit 4t' 1' F *F . -

F . '.. - - 'F F F t F F - . F,.

• F F

- JFFFJ - - F, '* F., ..--

F

• F 4 F, F,

F.F. F " , F- F -

- "F, F:?; ..,1F.r,' .F *,1.

1

-. j F

- '. - - F - F FF a F ' "F" - * 'F

'"'F ' F - 'F itt -

F, F" F,- ;-r--, 'V. .

', ' F IF F-, F FTF ' - *'I - F F F. F .1

-. 'F - 1/2.

HS1

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORY:

HS2 Control Room Evacuation Has Been Initiated and Plant Control Cannot Be Established EVENT TYPE: Control Room Evacuation OPERATING, MODE APPLICABILITY: All EAL THRESHOLD VALUE: '

The following conditions exist:

1) Control room evacuation has been initiated.

AND

2) Control of the plant cannot be established per AOP 915 within 20 minutes.

DAEC EAL INFORMATION:.'.

There is no significant* deviation from the generic EAL. :The applicable procedure for control room*

evacuation at DAEC is AOP 915. Based on the results of the analysis described below, DAEC uses 20 minutes as the site-specific time limit for establishing control of the plant. DAEC has satellite panels associated with the remote shutdown panel at various locations through out the plant. Control of the plant from outside the control room is assumed when the controls are transferred to remote shutdown panel 1C388 in accordance with AOP 915.

The EC/OSM is expected to make a reasonable, informed judgment within the 20 minute time limit that control of the plantfrom the remote shutdown panel has been established. The intent of the EAL is that control of important plant equipment and knowledge of important plant parameters has been achieved in a timely manner. Primary emphasis should be placed on those components and instruments that provide protection of and information about safety functions. At a minimum, consistent with the Appendix R safe shutdown analysis described above, these safety functions include reactivity control, maintaining reactor water level, and decay heat removal.

General Electric performed analyses to demonstrate compliance with the requirements of 10 CFR 50 Appendix R for DAEC. The evaluation of Reactor Coolant Inventory was performed using the GE evaluation model (SAFE). The SAFE code determines if the reactor coolant inventory is above the TAF during the safe shutdown operation. If core uncovery occurs, the fuel clad integrity evaluation is performed by determining the duration of the core uncovery and the resulting peak cladding temperature (PCT). The PCT calculations were performed by incorporating the SAFE output into the Core Heatup Analysis code (CHASTE). The details of these calculations are provided in Section 4 of the final report for DAEC HS2

,*IA * , "- *EBD-H

.~~Rev. - 4.

HAZARDS AND OTHER CONDITIONS FFECTING PLANT SAFETY Appendix'R a "ialys's'S"fe Shutd Appe..di. R Analyses foiIuiaiii&'Afiold Energy Cefiter";"MDE-44 036) . " '

The required analyses include evaluation of the safe shutdown capabilityofth6 rem6te shutdown system for various control room fire events assuming: (1) no spurious operation of equipment, (2) spurious operation of a safety-relief valve (SRV) for 20 minutes, (3) spurious operation bf a SRV for 10 minutes; and (4) spurious leakage from a one-inch line. The analyses show that the worst case spurious operation of SRV or isolation valves on a one-inch liquid line (high-low pressure interface) will not affect the safe'shutdown ability of the remote shutdown system for DAEC in case of a fire requiring control room evacuation before the identified time limit for.the necessary operator, actions at the auxiliary .shutdown panels. -ForTthe limiting cases of worst case spurious leakage from a one-inch line and spurious operation of a SRV,,operator control within 20 minutes would not impact the integrity of the fuel clad, the reactor, pressure vessel, and the primary containment.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 915, Shutdown Outside Control Room ,

2. General Electric Report MDE-44-0386,Safe Shutdown Appendix JAnalysisfor DAEC, March 1986" ,

3. UFSAR Section 6.4, Habitability Systems.

K 4. NEI Methodologyfor Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999 ":

fl ," f ,. -- S ,,

- -HS

EBD-H IHAZARDS _1 p"

AND OTHER COND~ITIONS A*FFECTING PLANT SAFETiY CATEGORYI -PAGE 26 of29 Rev. 4"

. ... L ,26 of29 :

HS3 Other Conditions ExistingWhich in the Judgment of the <EC/OSM> Warrant Declaration of Site Area Emergency EVENT TYPE: EC/OSM Judgment' OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

1. Other conditions exist which in the Judgment of the Emergency Director indicate actual or likely major' failures of plant functions needed for protection of the public.

DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL.

Per EPIP 2.5, the Emergency C"oordmatoi/Operations Shift Manager (EC/OSM) is the title for the emergency director function at DAEC: The EAL 'addresses conditions'that fall :under'the Site Area Emergency classification description contained in NUREG-0654, Appendix I.,

Events are in process or have.occurred which ijivolve actual or likely major failures of plant functions needed for protection of the public, Any releases are not expected to, exceed EPA Protective Action Guidelines beyond the site boundary but ?ould b.e exceeded onsite. ,

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control.Room Emergency Response Operation,

2. NUREG-0654/FEMA-REP-1, Criteria for Preparation and Evaluation-of Radiological Emergency Response Plans and Preparednessin Support of Nuclear Power Plants, Revision 1, October 1980, Appendix 1 HS3

EBD-H Rev.24 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORY PAGE 27 of 29 HG1 Security EventnResulting in Los§ Of Ability to Reach and Maintain Cold Shutdown '-4. " '

I EVENT TYPE: Security

• q4.44.4*.. 4 494 OPERATING MODE APPLICABILITY: All

'4 .4 444 EAL THRESHOLD VALUE:

4 4 - 4.

One of the following: - 444 -- .44.

44 "4 4 144' 4.

44 4 4

1. Loss of physical control of the control -roomdue to security' event. -'4 . '

OR

2. Loss of physical control of the remote shutdown capabiliti due to secirity'event.'

" IA TA h "T W 'I T 'T . ". ,. .-" 4 4' * .'::* 4::

mJkn%- rp.,

.t/ Inr11 U M* Ird1,1"Nll

' 4

- . .4.5 This EAL is an escalation of the SITE AREA EMERGENCY, HS1 declaration for a hostile force intrusion of a Vital Area taldni pj hysi6al control' of either the Contirol Room OR taking over The remote" shutdown capabilities which results hi the loss of physical conAr6l-of the, faility.'Tis'also includes areas where any switches that transfer control of safe shutdown equipmlrefit to outside the control room are located.

REFERENCES:

- . . ; --. " ' " "" 4"

1. Abnormal Operating Procedure (AOP) 914, Security EV'ents - -" .

2. UFSAR6Section'64;,Habitability Sysiem` .... .

2.U .SA ~ cioa ' *'4

"  : 4*'

, * "4

,, *÷. 4*

'4 14.4 4- 4 4

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EA BASES D ME EBD-H HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY CATEGORY-] I PAGE 28 of 29.

HG2 Other Conditions Existing Which in the Judgment of the EC/OSM Warrant Declaration of General Emergency EVENT TYPE: EC/OSM Judgment OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

Other conditions exist which in the Judgment of the Emergency Director indicate:

1) Actual or imminent substantial core degradation with potential for loss of containment OR

2) There is a potential for uncontrolled radionuclide releases. These releases can reasonably be expected to exceed EPA PAG plume exposure levels outside the site boundary.

DAEC EAL INFORMATION:

Per EPIP 2.5, the Emergency Coordmiator/Operations Shift Manager (EC/OSM) is the title for the emergency director function at DAEC ,

GENERAL EMERGENCY - 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.

IMMINENT - Mitigation actions have been ineffective and trended information indicates that the event or condition will occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

POTENTIAL - Mitigation actions are not effective and trended information indicates that the parameters are outside desirable bands and not stable or improving.

This Emergency Action Level allows for classification of events which in the judgment of the Emergency Director warrant the GENERAL EMERGENCY classification but do not fit into any other GENERAL EMERGENCY criteria. Emergency Director judgment is to be based on known conditions and the expected response to mitigating activities within a short time period arbitrarily set at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Classification of a GENERAL EMERGENCY is not to be delayed pending an extended evaluation of possiibilities and probabilities. If time allows and the offsite response organizations are active, consultation with the effected state and the NRC is prudent prior to classification.

HG2

EBD-H HAZARDS AND OTHER6CNDITIONS AFFECTING PLANT SAFETY CATEGORY

,Rev. 44 I "PAGE29 of 29:

4 J 144 -. -

REFERENCES:

1. Emergency Plan Implementing Procedure (EPIP) 2.5, Control Room Emergency Response Operation

2. NUREG-0654/FEMA-REP-1, Criteriafor Preparation and Evaluation.of',Radiological Emergency RespohsjePlaws and Preparedness in Support of Nuclear Power Plants, R~vision .1, Ociober 1980, Appendix I , ,, ,

3. NEI Methodolojyf i Development of Emergency Action Levels NUMARC/NESP-007 Revision"4, May 1999 4 444 41 4 4 4 4 4 7 4 q44,,44 -: .. 4** 4; 4. r 44474444 4, 444*

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- 4 44 4 HG2

-DAEC PROCEDURE WORK REQUEST (PWR)

PWR No.' 17582 Page 1 of I PROCEDURE QUALIFICATION IMPACT I NOTIFICATION REviEW (

Evaluate this procedure revision according to the criteria listed below, then indicate which of the following options is applicable:

QUALIFICATION IMPACT REVIEW ...... ] (TMAR No: ) NOTIFICATION ...... NEITHER ......... n-Procedure Owner - . Date:,

QUALIFICATION IMPACT?

If any of the following questions can be answered "yes", a Qualification Impact Review i"s're-quired.

1. Does this procedure revision incorporate the use of-equipment with new or different functions that require new or greater skill(s) by the user or b,' support personnel?

2. Does this procedure revision add to or change the sequence of steps ina Way that could have a significant impact on either plant operation or personnel safety?. .. -..

3. Is training required by regulation, external agency commitment, or.managemept direction?

If any of the above questions were answered with a 'yese, mark the QUALIFICATION :MPACT REVIEW block above and submit a copy of each of the following items to the Training Department for evaluation:

• a copy of the new revision of the procedure and associated PWR(s).

• a properly-completed Training Management Action Request (TMAR).

PROCEDURE NOTIFICATION?*' .. --

If either of the following questions can be answered "yes", notificationu of th4 groups indicated on the list below is required.

1. Does this procedure revision add to or change personnel responsibilities beyond what is currently required?

2. Is a general awareness of this procedure/procedure revision necessary to ensure compliance?

If either of the above questions were answered with a "yes", mark the NOTIFICATION block above and indicate by checking as appropriate on the list below which DAEC departiment(s) is/are to be notified of ihe'riew procedure revision:

DAEC OPERATIONS ENGINEERING TRAINING Safety ................E] Operations. ................. fProgram Engineenng .s...... ,, E] Operatibns Training ............ f Site Wide .......................... Other ROsI*ROs ;. System Engineering '..:...... D Technical Training .............. L

-El El- Project Engineering .......... [] Administrative .................

E] Emergency Planning Ii El MAINTENANCE OPERATIOItS ,LSUPPORT-, RAIDPROTECTION REG PERFORMANCE Electrical Maintenance.,-] Reactor Eng .... ..... ..... ] Radwaste ............... ] Licensing ....................... l Mecharical Maintenance.. [] Procedures ..... ..................... ] Health Physics ................... [: Quality Assurance . ........... []

I&C Maintenance ............. E] STAs ...... .............. D Chemistry ......................... [] Security .................... .

Metrology (M&TE) ............ Decontamination ................ LI ..jil oc ..... :..........................] Environmental ............. [1...

Civil .................... ]

Work Management ........ .. it,,"

OTHER (list) BUSINESS UNIT Business Unit ..................... F1 Nl- --DF1 NEITHER?

Based on the above cnteria, the new procedure revision requires neither a QUALIFICATION IMPACT REVIEW nor NOTIFICATION I _

NG-206K Rev. 0

EBD-S

--SYSTEM MALFUNCTION CATEGORY -~ .4 - '. Rev. 3 l!"',

Page 1 of 41 L Effective Date:-009 o;

ý II.

- 44 44'4' -

444.

.44.44 -4 Date: .. //

inaepeno PROCýEDIJRE APPROVAL .

I am responsible for the technical bofitent 6f this procedure a'nd for obtaining the necessary approval froni the Siatl, arid Count E'm'erge cyM nagem~entofialprr to implemren'tation. , .. 4.-"-.

Docume~ntati'on of State and C6 Wnt"YErn'frg'en~cy M'n~a'g'meht 'a'rovali s via "

N E Ob4 P-4 -4 030.4,4 *44.'

Approved Iby: ~---U Date2 Managier,'-Emergiency Planningi -

V ý_ - . -. 1

- 4 4 4444 44 44 4 -. . A 4 444 4.4

.44 5 - *44

Table of Contents SUI' Loss of All Offsite Power to Essential Busses forGreater,,Than-15. Minutes ..... S-4 SU2 - Inability to Reach RequiredShutdown Within Technical Specification Limits .... S-6-ý SU3 - Unplanned Loss of All Safety System Annunciation or Indication in the Control Room for Greater Than 15 Minutes -.................... S-7 SU4 - Fuel Clad Degradation ..................................... S-9 SU5 - RC S Leakage .................................................................................................. S-13 SU6 - Unplanned Loss of Alt Onsite, orOffsite Communications Capabilities .......... S-15 SU7- Unplanned Lbssof Required DC Power During Cold Shiutdown or Refuel Mode For Greater Than 15 Minutes .......................................................................... S-17 SA1 - Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses During Cold Conditions ................................................................................... S-19 SA2 - 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 ................ S-21 SA3 - Inability to Maintain Plant in Cold Shutdown .................................................... S-23 SA4 - 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 ............................... S-25 SA5 - 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; .............................................................................. S-27 SSI - Loss of All Offsite Power and Loss of All Onsite AC Power to Essential BussesS-28 SS2 - Failure of Reactor Protection System Instrumentation to Complete or Initiate on Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was NOT Successful ............................. S-29

SS3 - Loss of AllVital DCoPower. e..r............................................S-31 SS4 - Complete Loss ofFunction Needed to Achieve or Maintain Hot Shutdown .... S-32 SS5 - Loss of Water Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reactor Vessel. . "........................................................................................ S-34 SS6 - Inability to Monitor a Significant Transient in Progress.: .S-36 SGI - Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Pow er .............................................................................. ,,...... ........................ S-38 I SG2 - Failure of the Reactor Protection System to Complete an Automatic Scram and Manual Scram was NOTSuccessfui arid Tihere'i*'Ihdi6*ation of an Extreme' -

Challenge to the Ability to Cool the Core ........................................................ S-40 4 - 5 *-

444*** 4-j -

SUI Loss of All Offsite Power to Essential Busses for Greater ThaW -*

15 Minutes EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: All EAL THRESHOLD VALUE:

The following conditions exist:

1. Unplanned loss of power to both Startup (1X3) and Standby (1X4) transformers is expected to last for greater than 15 minutes.

AND

2. Emergency Busses 1A3 and 1A4 are powered by their respective Standby Diesel Generators.

DAEC EAL INFORMATION:

UNPLANNED - The loss of power is not the result of a planned evolution.

This event is a precursor of a more serious Station Blackout condition and is thus considered as a potential degradation of the level of safety of the plant. It is possible to be operating within Technical Specification LCO Action Statement time limits and make a declaration of an Unusual Event in accordance with this EAL.

The intent of this EAL is to declare an UNUSUAL EVENT when offsite power has been lost and at least one of the emergency diesel generators has successfully started and energized at least one ECCS bus.

Sul

REFERENCES:

1. Abnormal Operating Procedure (AOP) 301, Loss of Essehtial Electrical Power

2. UFSAR Section 8.2, Offsite Power System ,

3. NEI Methodology for Deleloprnientof EmergencyAction Levels NUMARCINESP-007 Revision 4, May 1999 , ":

1.'.*, - ' .A " ,, , **'I ,;:,'. . " , .- - .,* , .'L; " : . ' - '

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• 1 SU1-

EBD-S

-SYSTEM MALFUNCTION CATEGORY', Rev. 3 "Page6 of 41 SU2 Inability to Reach, Required-Shutdown Within -Technical Specification Limits EVENT TYPE: Tech. Spec. LCO Action Statement TimeLimits. Expired OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown' EAL THRESHOLD VALUE:

The following conditions exist:

1. Plant is NOT brought to, required operating mode within the Technical Specifications LCO Action Statement.-,lii'e.' .

DAEC EAL INFORMATION:

Limiting Conditions for Operations (LC1O) require the pla'nt to be brought to a specific condition when an LCO has been entered. Depending on the circumstances this may or may, not be an emergency or, a.precursor to a more serious, event. in any case when a plant initiates a shutdown due to. having entered an LCO, action statement a one hour report must be made under 10CFR50.72(b) non-emergency events. The plant is within its safety envelope when being shutdown within the allowable action statement time of a Technical Specification. An immediate classification of UNUSUAL EVENT should be made when the plant is.,,NOT brought to the required mode within the allowable action statement time of any Technical 'Specification, LCO. Declaration, is based on the time at which the LCO Action ,Statement specified time -period elapses and is NOT, related to how long a condition may .have existed., ,

REFERENCES:

1. DAEC Technical:Specifications: , .

2. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May-1999, - .. I, 1

SU2

EBD-S SYSTEM MALFUNCTION CATEGORY`:_..',%; Rev. 3 Page 7 of 41 SU3 Unplanned Loss of All ,Safety System Annunciation'br Indication in the Control R6m for Greater Than 15Minutes, EVENT TYPE: InstrumentationmComniunication OPERATING MODE APPLICABILITY:.,Run, Startup, Hot Shutdown. ,

EAL THRESHOLD VALUE:

The following conditions exist:'

1. Unplanned ioss-of mdsf br all - C03, 1C4O and 1C05 Annunciator or indicators,;

associated with Critical Safety Functions for greater than 15 minutes.

AND .'-

K. 2. Compensatory non-alarming indicatioris-eare available.,

3.` In the opinion of the' Operations` Shift Mýnager lhid16ss of afiuiuhciat6rs-:or indid6tor" requires increased surveillance to'safely'operate theni't.-

t -

DAECEAL'INFORMATION: - -.2 " "

Control room panels lC03, IC04, `,and 1C05 contatin'etarrn'nunciators associated with safety systers :at'DAEC., Theiefoi'e, the-D'A'EC EAL'2a'ddrsse§§s un-planied-loss bf most" annunciators 6ro th&e'e pariels:?'Conipefnsatbrynbnh-alaýrnng indicatibns-includes the6'lant' process computer,"-ISPDS,'* plant riec6ders, or-plant 'ihstrnrhent*'"displays in *the 'control, room. Unplanned loss of annunciators or indicators excludes scheduled rnaintenance and testing activities.

Under the conditions of concern , entry into AOP 302.2, 'Loss -of .Alarm Panel Power,'

would be made. Thf procedure: requis b lerting oberat6rý-on'6hift to thi nature 6f the lost annunciation. It further requires that operators be attendant and responsive to abnormal indications that relate to those systems and components that have lost annunciation. Therefore, the generic criterion related to specific opinion of the Operations Shift Manager that additional operating personnel will be required to safely operate the unit is not included in the DAEC EAL because the concern is addressed by the AOP.

SU3

MOST - 75% of safety system annunciators or indicators are lost OR a significant risk that a degraded plant condition could go undetected exists.[ THe use arid definition of.

MOST is not intended to require a detailed count of lost annunciators or indicators but should be used as a guide to assess the ability to m6nifor the operation of the plant.

Unplanned loss of critical safety function indicators (i.e., EOP/EAL parameters) for greater than 15 minutes may preclude operators from taking actions to mitigate a transient.

Annunciators on 1C03, 1C04, and ICO5 share a common power supply from 125 VDC Division I that is fed through circuit breaker 1D13. ,. - .

Indications of loss of annunciators associated with safety systems include:

• 125 VDC charger, battery, or system annunciators oncontrol room panel 1C08

• Loss of "sealed in" annunciators at affected panels

& Failure of affected annunciator panels shiftily testing by plant operators

° Expected alarms are not received . ' '

0 Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, 1C04, andlC05),

1- -

REFERENCES:

1. Operating Instruction (01) No.1317.2,Annunciator.System, .,

2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power .

3. Abnormal Operating Procedure (AOP).302.2, Loss of Alarm Panel Power

3. NEI Methodology for Developmernt of Emergency Action Levels NUMARCIVESP-O07,

'Revision 4, May 1999 SU3

SU4 Fuel e Clad a Degradation ra . , +" .-...

.... +

EVENT T  : CoolantActivity '- " - - 4 OPERATING MODE APPLICABILT'":All ' '- P--

EALTHRESHOLD VALUE:

One of the folloiing: - .' ' N' -  : . . *. . -- -"." - .,

.... 4 ' 4 + +

1. Valid pretreat radiation monitor (RM-41 04) reading greater than 4E+3 hiR/hr 4 -. .,., . , " ,; +-; , -

2.--Reactor'f- Coolant 'ýsarmple" activity -Value indicin'g. 0greater"+than" 1.2 Cilml "dose equivalent -' . "

1-131. . ., * .-.. .

4, + " .: '..., - -,., ; + r'" .....

DAEC EAL INFORMATION:-

-*- . * '4

• *- . . ,4. . * - * .4 + ;+ 4 . "

• There are no significant deviations from the generic EAL.s'IThesie:.EALs-are precursorsof miore serious fuel clad degrada'tidn and are thus' consideredd indicating a potential degradation of the level of safet} of the plant. ,to Thu. , it is'possible be ophrting:within Technical Specification LCO Action,Statement time'limits for iodine spikes and make -a declarationof an :UnusualEvent.:; DAEC, mode -applicability-f6r these EALs are consistent with the Tech Specs.'"+ '. - , , ,

SEAL '~laddressesvalid pretreat rad :monitor exceeding (RM-4104) aboge-.4E+3 mR/hr."

"+The calculation supporting this value Iis described below:. Valid means.:that the pretreat 4rad monitor reading is determined to be operable in accorda'nce with the Technical spec~ificati0ns6r °dS' be166n Verified by'.7t*er inepered-eridnt-rm6-ehod.-ý7ucJha-sa inflidations'

• lispl51ay ed on the6 6toItrlr-nels, reports fr6irffplant persronnel,Ior _o0661nt samplinrg resu lts.

This reading Would bebidisplayed on -Control Room panels1 C-02 and C-10 on recorder" RR-4104.' 4- ............. '-..--. ----- 4..

As specified in- the eneric methodology, DAEC EAL'-2 addresses -coolant samiles "exceeding technical specification 3.4.6,-coolant activity less than-or equal to 1.2 ciCiml dose equivalent 1-131. .,.-+

-. , .- 4\ , . + - .- '4 -.+ , 4' , .

SU4

l7A*A,-i U EBD-S

]L)

SYSTEM MALFUNCTION CATEGORY,,, Rev. 3 Page 10 of 41 Radiological Engineering Calculation 94-014A and, UFSAR Table 15.4-1 were; reviewed to' determine a suitable- EAL threshold for the pretreat rad monitor reading corresponding to,'

the Tech Spec 3.4.6 coolant activity limit of 1.2 tiCi/ml of dose,equivalent 1-131. Using the condenser noble gas source term for the control rod drop accident of 2.38 E +06 Curies shown on UFSAR Table 15.4-14and the condenser' free volume of, 55,000 cubic feet,- an:

initial noble gas concentration in the condenser offgas line is determined. Because the offgas' flow rate is very small (about 50 standard cubic feet per minute) compared to the total condenser free volume, dilution of the condenser noble, gas concentration due to offgas flow is not considered in thecalculation shown below. Decrease in the noble gas source term due to decay of short-lived noble gas radioisotopes and offgas flow dilution

'effects are addressed, by rounding down the value calculated as shown~below,.:

Calculation 94-014A used an exposure rate, method based on using a source term consisting of a defined-mixture ,of noble gases and iodine from the, control rod drop

'accident as described, in the DAEC UFSAR, Section 15.4. The calculation, assumed that the activity is released instantly and immediately reached in equilibrium with the reactor coolant inventory. Using this calculation, using dose correction factors (DCFs) for child thyroid dose from Reg. Guide-l.109, and adjusting for the specific gravity, (0.736),of.

saturated water at. 1050 psia (fluid conditions assumed in,the calculation) to adjust for standard conditions;,_the, 1-131, dose equivalent (in, units qf pCi/ml assuming 1 cc equals-1:.

ml),, is determined -,for,,this. event. ,.:This, result is- then, linearly scaled. for, rad monitor, readings corresponding. to the. Tech Spec 3.4.6 allowable primary, coolant activity of 1.2 LiCVmI 1-131, dosa; equivalent, 1i.e;, the relative, mixture.,of noble'.gases, and iodine is assumed to remain constant. 1-129 is ignored because it'has no effect on the calculation result.... . *. . ....

Isotope DCF. c, -. Concentration Correction, Factor.--, 1-131 DEQ (jtCi/cc)

(mrem/pci):---- (pCi/cc) - [DCFSOTOPE I DCFi. 13,]/;

• ~0,736, 1-131.. 4.39 E-03 -1.6 Et01 - . -':,1.4 E+00,. 2.2'E+01 1-132 5.23E-05. 2.2E+01 1.6 E-0 *-01 1-133 1.04-E-03 - 3.J1, E+01.- 3.2 E-01 1.0 E+01 1-134 1.37 E-05 3.4 E+01 4.2 E-03 1.4 E-01 1-135 2.14 E-04 2.9 E+01 6.6 E-02 1.9 E+00 TOTAL -- 3.4 E+01 Therefore, for this event, a coolant activity of 34 pCilcc 1-131 dose equivalent is calculated. Scaling the results for 1.2 pCi/cc 1-131 dose equivalent, a suitable condenser SU4

V '

EBD-S SYSTEM MALFUNCTION CATEGORY .') Q . -' Rev. 3 Page 11 of 41 source term and corresponding initial concentration in the offgas flow is then 'dete'rmined.

This is then converted to -a pretreat rad monitor- reading by use of the monitor efficiency factor.  ;'~ .'j ~ ' *s ':

., ". Pretreat Rad Monitor (RM-4104) Reading.,-

• NG concentrationcdad damage = NG conc6ntrationho"DDRoP*X [1:2 [tCi/cc /34"tCi/6ci]

= 1[2.38 E +6 Ci "x I E+6 -PCi ICi]/ [5.5 E+4 ft3 x 2.83 E+4 cc/ft3] X [1.2 Ci/V66 /34 pCi/cc]

= 1529 [.Ci x/'0.0353 =-54.0 ptCi/cc Pretreat rad monitor reading =NG concentration ,X Rad monitor efficiency "Rad,monitorefflciency = 89.2Rhir i/c thdre re Rad moit' rgei '89.2X

' Pretreat rad'monitor reading *=`0.2:X 54:0 '=4800-rmR/hr '

To account'for isotopic decay and dilution "effectsof offgas flo~v, round down to 4E+03 S. .... . . .... . . m P; hr , -. ,

." . .,l* I . -,'

The balculation results were also re1iewedAto determhine if suitable'vaIues for the'main steam line (MSL)'radiation monitors c'duld'be developed. 'As showri above,.the rod drop accident corresponds to' cbolant ctivity of '34:,iCi/cc 1-131" dose iequiiva-len't. "As determined by the reference -caicu^lati6n, 'this -corresponos to a MSL radiation monitor reading of about 5.7 R*hr." Scaling~the results for-1.2 pýCi/ml 1-131 'doge equivalent:'

MSL Reading Corresponding to 1.2 pCilml 1-131 dose equivalent

((1.2 pCiI66]I (34 Ci/ccj]'X 5.7 Pr/h--` 0.2 R/ir~ 200 m-R/hr 200 mR/hr is at the Jower-end d.'.The normal MSL monitor readings during full' power.

.Because this value is. not .distinguishable,. and -'hydrogen water chemistry system malfunctions that result in increased production of N-16 can -alsoresult in increased main steam line radiation levels, it is pot appropriate at DAEC to use thle nriain steamn line monitor readings. "".

SU4

REFERENCES:

1. Abnormal Operating Procedure (AOP) 672.2, Offgas Radiation/Reactor Coolant High, Activity

2. Technical Specification 3..6-, Coola it Chemistry,

3. Radiological Engineering Calculation No. 94-014A, Main Steam Line Radiation Monitor Setpoint Calculation, August 29, 1994

4. Surveillance Test Procedure (STP) No. 3.4.6-01, Reactor Coolant Gamma and Iodine Activity

5. Annunciator Response Procedure (ARP) IC03A, Reactor and Containment Cooling and Isolation

6. Annunciator Response Procedure (ARP) lC05B,' R6actor'C6-nt' "'[

7. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999 I

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2 N A, *Z Nt.

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SU5 RCS.Leakage EVENTTYPE: Coolant Leak " . '

OPERATING MODE APPLICABILITY`:'Run,,Startup, Hot Shutdown, Cold Shutdown EAL THRESHOLD VALUE: "

• Oneof the following:- .

ggpente I~Itth.

1. Unidentified or , n a than 10.gprn. "

'O R '" ' .,. - . .. '" ' " " "";,,::t ,, '

2. Identified leakage greater than 25 gpm.

OR

3. Valid indication of Main Steamline Brea-k",

DAEC EAL INFORMATION: .

EAL Threshold Values 1 and 2 are precursorsof more serious RCS barrier-challenges and are thus considered as a potential degradation of 'the level of safety of the plant.

Thus, it is possible to be operating within Technical Specification LCO Action Statement time limits and make a declarationof an Unusual Event in accordance-with these EALs.

Credit for the action statement time limit should'only be:given when leakage "exc6eeds technical specification limits but has not yet exceeded the Unusual Event EAL~thresholds described above. In addition, indication of main steam line break has been added here as discussed in NUMARC Methodology for Development of Emergency Action Levels NUMARC1VESP-007 Revision 2 Questions and Answers, June 1993, Fission Product Barrier-BWR section. This was in response to question 4 which states that the main steam line break with isolation can be classified under System Malfunctions.

Valid means that the reading is from instrumentation determined to be operable in accordance with the Technical Specifications or has been verified by other independent methods such as indications displayed on the control panels, reports from plant personnel, or radiological survey results.

SU5

The DAEC Tech Spec Section 3:4.4 coolant system leakage LCO liimits- are: (1) *! 5 gpm unidentified leakage, (2) < 25 gpm total leakage averaged over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, and (3) _<2 gpm increase in unidentified leakage within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />' period in Mode 1. Total leakage is defined:as the sum of identified and unidentified leakage.

DAEC EAL Threshold Value 1 uses the generic value of 10 GPM for unidentified leakage or pressure boundary leakage. The 10 gpm value for the unidentified or pressure boundary leakage was selected as it is observable with normal control room indications.

DAEC EALThreshold Value 2 uses identified leakage set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressure boundary leakage. .. ,

REFERENCES:

1. Technical Specification 3.4.4, Coolant Leakage

2. Surveillance Test Procedure No. (STP) 3.0.0.0-01, Reactor Coolant System Leak Rate Calculation

3. Operating Instruction No. (01) 920, Drywell Sump System

4. Alarm Response Procedure (ARP) 1C04B, Reactor Water Cleanup and Recirculation

5. Alarm Response Procedure (ARP) 1C04C, Reactor Water Cleanup and Recirculation' 6., UFSAR Section 5.2.5,.,Detection; of Leakage, through ,Reactor Coolant Pressure,,

Boundary

7. UFSAR Section 15.6.6, Loss-of-Coolant-Accident f: , .

8. NEI Methodology for Development of Emergency Action.Levels NUMARC/NESP-007 Revision 4, May 1999 -'  :,,, ,T 1%

SU5

PIage 15of41 SU6 _Unplanned.Loss-of All Onsite or Offsite Communications Capabilities " - " 4 '4 EVENT TYPE lnstrurmentation/Comn-munication< 4 ...

OPERATING MODE APPLICABILITY:'A l " "

EAL THRESHOLD VALUE: ,-

' ,., '" r .. , *, .':- , 4 -* ', ,.

One-of the'followkinggroups of communication 'losses: "

1. Loss of ALL of the following '6nsite' cormmunicati6n' 6apabilities'affecting the'ability to' peif6rm'routine operation: "

" Plant Operations Radio System ' . "

", :Plant Paging System ' ,

;n-plantTelephones ",:, - - -..

Sound Power Tel6phohies , , , . . 4 . 4 2.-Loss'of ALL of thiefollowing offsite communiCations *paebility SAll telephone lines (commercial)  :-4 C. .. .

"4 Microwav6"PhoneSystehi',; ,

• FTS-2000 phone system (ENS & HPN)

• Cellular Phones DAEC EAL INFORMATION:

There is no significant deviation from the generic EAL. The communications methods used at DAEC are described in the Emergency Plan. In-plant and external agency telephone communication methods include PABX lines, direct-ring lines, and NRC telephones which are extensions for the Emergency Notification System. There is also a microwave system to provide backup emergency telephone communications.

The availability of one method of ordinary offsite communication is sufficient to inform Sjstate and local authorities of plant problems. This EAL is intended to be used only when SU6

-.SYSTEM MALFUNCTIOI C*ATEGORY' " Rev. 3 Page 16 of 41 extraordinarymeans (relaying of information from radio trahsmissions, individuals beinig sent to offsite locations,etc.) arebeing utilized to make communicationspossible.

The DAEC plant operations radio system is a UHF system with. consoles located in the Control Room, Technical Support Center, Operational Support Cente'r, 'and the Central Alarm Station. Hand-held, transceivers are Iused in., this, system, to provide simplex.

communications within the plant and onsite. The DAEC Radiological Survey Radio System is an 800 MHz trunked/conventional repeater system, that provides base-to-'

portable communications throughout the DAEC EPZ. A secondary high-band system provides back-up capability for the 800 MHz radio. -Consoles are located in theTechnical Support Center and the Emergency Operation's Facility at the IES Tower. The DAEC Security (backup,-radiological- survey) Radio System provides,* base-to-portable. security communication within the plant and with the Linn County SheriffsOffice using a mobile relay (repeater) type base station and two VHF frequencies. Control consoles are located in the Secondary Alarm Station, Central Alarm Station, Security Control Point, Technical Support Center, and Emergency Operations Facility. The DAEC also has a base'station licensed for-operation in the Police- Radio-Service. on the law enforcement state-wide, point-to-point VHF frequency. The transmitter and one control console are located at the Secondary Alarm Station and in the Central Alarm Station. This station is for communications with Iowa Department of Public Safety radio' "station, -Linn County,,

Sheriffs office, and the Benton County Sheriffs office. This point-to-point channel is also used by- the Linn County Emergency Management and other public-safety, organizations, throughout the state of Iowa., . ,  ;-; ,- -

REFERENCES:

1. Emergency Plan, Section F, Emergency Communications

2. NEI Methodologyfor Development of Emergency.Action Levels NUMARC/NESP-007 Revision 4, May 1999 -: - ', . ,

SU6

.4-'.

EBD-S

.-,-SYSTEM MALFUNCTION CATEGORY,-- ,,-.. Re.

-.Page 17 of 41 t

SU7- Unplanned Losspof Required DC-Power During Cold Shutdown or Refuel Mode For Greater Than 15-Minutes EVENT TYPE: Loss of Power " ,- ,' , ...

OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel.- " -"

EAL THRESHOLD VALUE: -.. , .. ,

The following conditions exist: - ,- .

1. Unplanned'loss of Divisiontl (1D1) and Division 11 (1D2)-125 VDC buskes bas6d on bus'

-voltag'6.indicati~ris .ý- - ~- '-  ;

AND . "

2. Failure to iestore power to at leastone req'uired DC buis within 15 minuies-from time of DAEC EAL INFORMATION:. ' " - - " ,

- . - . 'I I --  : j There is no significant deviation from the generic EALI.; lUnplannied loss of.DiV. I and Div.'

II. 125 VDC busses excludes scheduled maintenance and testing' activities. Underi the conditions of concern, AOP 302.1, Loss of 125 VDC Power, would be entered. The DAEC EAL's address the loss of both divisions of the 125 VDC systems consistent'with AOP 302.1.

The 125 VDC-system is 'divided into two independent divisions - Division I (1D1) and Division II (11D2) - each with separate AC and DC (battery) power supplies. Loss of both 125 VDC Divisions could compromise the ability to monitor and control the removal of decay heat during cold shutdown or refueling operations. These EAL's are intended to be anticipatory in as much as the operating crew may not have necessary indication and control of equipment needed to respond to the loss. If this loss results in the inability to maintain cold shutdown, the escalation to an Alert will be per SA3 "Reactor Coolant temperature to exceed Technical Specification limit of 212 F or UNCONTROLLED temperature rise approaching the Technical Specification limit of 212".

SU7

EBD-S SYSTEM MALFUNCTION CATEGORY., ' ' Rev. 3 Page 18 of 41 Bus voltage is based on the minimum, bus-voltage necessary for-the operation of safety related equipment and.the loss may be indicated by, the illumination of annunciators "125 VDC System I Trouble' on IC08A A-9 andlor "125 VDC System II"Trouble"on IC08B A 4.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power

2. Abnormal Operating Procedure (AOP) 388, Loss of 250 VDC Power

3. Technical Specification 3.8, Electric Power Systems

4. UFSAR Section 8.3,,Onsite Power Systems

5. UFSAR Table 8.3-6, Plant Battery System -'DC Power, Instrumentation, and Control, Principle DC Loads (125V)

6. ARP 1C08AA-9

7. ARP IC08B A-4

4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999:...... ,,_,"

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- I I.

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SU7

- zA- A ESD,.M EBD-S SYSTEM MALFUNCTION CATEGORY , '., .- Rev. 3 Page 19 of 41 SAl Loss of All Offsite Power anhd Loss-of All Onsite AC Power to Essential Busses, During old Sh6tdown'Conditions-,

EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Cold Shutdown, Refuel, Defueled EAL THRESHOLD VALUE:

The following conditions exist: ' ,

I. Loss of power to both Startup (1X3) and Stanidby.,(lX4) transformers.

AND  ;....' ,

2. Failure of "AK Emergency Diesel Generator 1G-31 and tB" Emergency Diesel Generator to supply power to emergency busses 1A3 and 1A4.

AND- - a * *

3. Failure to restore power to at least one emergency bus, 1A3 or 1A4, within 15 minutes from the time of loss of both offsite and onsite AC power.

DAEC EAL INFORMATION:

Under the conditions of concern, entry into AOP 301.1, Station Blackout, would be made under Tab 1. Indications/alarms related to station blackout are displayed on control room panel 1C08 and are listed in the procedure under "Probable Indications."

The loss of both offsite and onsite AC power to the emergency buses when in Cold Shutdown, Refuel or Defueled modes, compromises safety systems required for decay heat removal and is a substantial degradation of the level of safety of the plant. An ALERT is declared in Cold Shutdown and Refueling modes due to the less severe threat to the protection of the health and safety of the public because of the much longer time available to restore power and decay heat removal systems.

15 minutes was selected to exclude transient or momentary power losses.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout KJ2. Abnormal Operating Procedure (AOP) 301, Loss of Essential Electrical Power SAl

I I I

3. Technical Specifications Section 3,8; Electrical Power Systems t.

4. NEI Methodology for Development of Emergency Action Levels,NUMARC/NESP-007 Revision 4, May 1999

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• j SA1

EBD-S

,SYSTEM MALFUNCTION CATEGORY. ,',,zRev.3 K ,Page 21 of 41 SA2ý Failure of ReactorProtection System Instrumentation to.

', Complete or Initiate an Automatic Reactor"Scram Once'a .

Reactor Protection System Setpoint Has Been, Exc6eeded and Manual Scram Was Successful, EVENT TYPE: RPS Failure - '

OPERATING MODE APPLICABILITY: Po Wer Op~ration, Startui" -"

EAL THRESHOLD VALUE"-'- '- .

The following conditions must exist to declare this EAL:

1. Auto Scram Failure - -

AND -.

2. Operator actions to reduce power are SUCCESSFUL as indicated by either:

a. ALL Rods Full-In, .

OR -'V

b. Reactor Shutdown Under All Conditions WithoutBoron, i-.

c. Reactor power below the APRM Downscale Alarm on ALL valid APRM instruments DAEc EALINFORMATION: ,' ,' ,

The condition of concem is failure of the Reactor Protection System (RPS) to scram the,,

reactor when a valid scram signal is present. 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 plant safety has been compromised and design limits of the fuel may have been exceed6d., -'

The EAL evaluation should -o5ciJr _after operatos -have taken actions from the 'main control room to insert a manual scram and reduce reactor power. Permissible actions iquickly fro'mthe main Control -room by oin-*hift" include all actions that can be pe'rf6rmred operators (e.g., use of the Manual Scram pushbuttons, ARI, placing the Mode Switch in" K Shutdown, individual scram test switches, etc.). It is not appropriate to delay the EAL SA2

EBD-S evaluation until -other time; consuming actions 'are 'completed. such as manual rod' insertion or completionof in-plant -EOP, Support Procedures for rod insertion (e.g.,

venting the over-piston areas of individual CRDs).

Operator actions are considered successful if any of the following results are achieved:,

" All control rods inserted to at least position 02 - this is defined in EOPs as the Maximum Subcritical Banked Withdrawal Position and is the lowest control 0rod position to which all control rods may be withdrawn in a bank and the reactor will, none the less remain shutdown uncder all conditions, irrespective of reactor coolant temperature and any boron which may have been injected into the RPV.

" Determination that the Reactor is "Shiutdown'unde6r ALL conditions without boron" this can be determined by relying on the Technical Specification demonstration of adequate shutdown margin:

- One control rod is out beyond position 00 AND

- All other control rods are at position 00 For other combinations of rod, pqaterns and boron concentration, reactor engineering will need to perform a shutdown margin calculation. .

" Reactor power is below the APRM Downscale Alarm Setpoint on ALL valid APRM instruments.

Note - If the mode sivitch,is in St~rtiýp 'andtherods are fully ihs6rted (i.e.,.'the reactoris shutdown) prior't6 the dutonatic signal failur6, -then declaiatioh'of-h'Alert'wbouldnot b6e required. In this case, tfdeeverht wduld"be eported uJnd&riO CFr-? 50.'72 (b)'(2) (I)-as a four hourreport. - , . ' - ' &

REFERENCES:

1. Integrated Plant Operating, Instruction (IPOI) No. 5, ReactorScram,

2. ATWS Emergency Operating Procedure (EOP) - RPV Control

3. Emergency Operating Procedure (EOP) 1 - RPV Control 4., NEI,'Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999 SA2

(

t, SA3 Inability to'Maintain Plant in Cold Shutdown -,'  :'"

EVENT TYPE: Inability to Maintain Shutdown' ConditioS:'"

OPERATING MODE APPLICABILITY: Cold Shutdowd ful' '

EAL THRESHOLD VALUE: . .

JJU Ol[I*

Tl 1UIIUWlI*y. .,,

1. Loss of Decay Heat Removal systems required to maintain Cold Shutdown.

^kl- , " , *"* -":**t-l} ,. *: t., -- '-I.- "...... - ,... . .

2..With CONTAINMENT CLOSURE not establishied, temperature conditions 'exist'that either: . - ". " '"

a. Cause reactor coolant temperature to exceed the Technical Specification limit of 212 OF.

OR "'-, 1' "" '

b. 2 Result in an UNCONTROLLED teinperatiki" rise approaching the'"

Technical Specification himit of212°F. ` .....

DAECAL INFORMATION: '"

Under,the conditions of, concern.for-EAL Threshold Value ,l,.AOQP. 149, Loss of Decay Heat, Removal, would, be entered, under, Tab;._.1,; Loss,-,,0f Shutdown ,Cooling.

Indic-ationrs/alarm's related t16 oss. 1*pf- shutdown. cooling are displayed on control room panels 1C03 and 1C05 and are listed in the procedure under "Probable Indications." The procedure requires that shutdown cooling be re-established.

The procedure provides curves of maximum water heat up rates which provide an upper bound of the heatup >until-an estimated time to -boil calculation carn be completed by Engineering.

The DAEC EAL is written to imply an RCS temperature rise above 212 OF that is not allowed by plant procedures. This corresponds to the inability *to maintain required temperature conditions for Cold Shutdown. "Uncontrolled" means that system K temperature increase is not the result of planned actions by the plant staff. The wording SA3

iSYSTEM MALFUNCTION CATEGORY. "  : *--" Rev. 3 i Page 24 of 41 is also intended to eliminate, minor; cooling interruptionis !occurring .'at th6. transition between Hot Shutdown and Cold, Shutdown or temperature changes that are permitted to occur during establishment of alternate core cooling so that an unnecessary, declaration, of an Alert does not occur. The uncontrolled temperature rise is necessary to preserve the anticipatory philosophy of NUREG-0654 for events starting from temperatures much lower than the cold shutdown temperature limit.

REFERENCES:

1. Abnormal Operating Procedure (AOP) 149, Loss of Decay Heat Removal

2. DAEC Technical Specifications I '

3. Surveillance Test Procedure (STP) 3.4.9-01, Heatup andCoold~wn Rate Log

4. NUREG 1449, Shutdown and Low-Power Operation at C nme-rcial Nuclear Power

. Plants in the United States, September 1993 , *. , '- t L 1 NEI Methodology for Devlelopment of ErnergencyAction Le',els NUMARC/VESP-007 Revision 4, May 1999_

- A.

SA3

4 S"-" 10 1 rIIVl IVI/LrUV1M.,I IIUI4*I ILIN- %, II I =%.DUV. ' 3

" "Page 25 of 41 SA4. Unplanned Loss -of Most or All Safety Sys~tem Annunciation or Indication in Control Room With Either (1) a Significant Transient inPro'gress',' or(2)'C6ompensatory Non-Alarming

..Indcatrsaeu . -. .,, , " ".

EVENT TYPE:. InstrumentationlCommunication -:  ;, "

OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL'THRESHOLD VALUE: ." "

The following conditions ex=it:, . . - * ,*.I , *. \ -. " ,

1. Unplanned loss of most or all 1C03, ;1C04 and -1C05?' Anndnciato~s -or',indicators.

associated with Critical Safety Functi6h s for gqeater.than 15 minutes.' .. * "

K> j AND 1.- In the opinion of the Operations Shift Manager, the loss of all"annunciat6rs 6r indlicators requires increased surveillance-to safely bperate the unit. -

AND

2. Either of the following conditions exist: " :'w ' -;

a. A significant plant transient in progress., -,. .- -

OR a a a a

b. Loss of all idicaton needed to monitor criticality, core heat removal, OR Fission Product Barrier status. -. .. ,- . ........

DAEC EAL INFORMATION: a a Control room panels 1C03, 1C04, and 1C05 contain the annunciators associated With safety systems at DAEC. Therefore, the DAEC EAL addresses,, unplanned loss of, annunciators on these panels. Compensatorynon-alarming indications includes the plant' process computer, SPDS,'-plant recorders, or plant'instnimentdisplays'in the -control room. Unplanned loss of annunciators or indicators excludes 'sch edu'maintenance, and testing activities. Significan't transient inclufdes' resplonse to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power

\I change, ECCS injections, or thermal power oscillations of 10% or greater.

SA4

EBD-S SYSTEM MALFUNCTION CATEGORY, - Rev. 3 Page 26 of 41 Under the coditiondt of concern, entry, into AOP 30.2, Los' of Alarrn Panel Power, would be made: The pr6cedurei- equires alerting oqperators' o'sihiftt6 thie nature of the lost annunciationi It further requires" that ' operators, be attendant- and responisive to abnormal indications that relate to those systems and components that have lost annunciation. Therefore, the generic criterion related to specific opinion of the Operations Shift Manager that additional operating personnel will be required to safely operate the unit is not included in the DAEC EAL because the concem' is" addressed by the'AOP.

MOST - 75% of safety system annunciators or indicators are lost OR a significant risk, that a degraded plant condition could go undetected exists. The use and definition of MOST is not intended to require a detailed count of lost annunciators or indicators but should be used as a guide to assess the ability to monitor t ibatiori f the. plant.

Unplanned loss of critical safety-function indicators (i.e., EOP/EAL parameters) for greater than 15 minutes may preclude operators from taking actions to mitigate a transient.

Annunciators on 1C03, 1C04, and,' 105 share, a common: powor, supply from .125 VDC, Division I that is fed through circuit breaker 1D13. Therefore, DAEC does not specify a loss of "most" annunciators as specified in the generic methodology..,.

Indications of loss of annunciators associated with safety,systems include: ,

125 VDCh carger, battery, or system annunciators on control room panel 1C08 Loss,of "sealed in", annunciators, at affected panels -i-I' . I Failure.of affected annunciator, panels, shiftily~testing by plant operators Expected alarms are not received,-:' ' -- .

Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC power to panels 1C03, 1C04; and 1C05). , 4', ,,X '.

REFERENCES:

1. Operating Instruction (01) No. 317.2 Annunciator System 2: "Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power

3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power

4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999 SA4

>C.SM T EBD-S

.. SYSTEM MALFUNCTION CATEGORY Rev. 3 "u Page 27 of 41 SA5 AC Power-Capability to Essential Busses Reduced to a Single power source for Greater Than 15 Minutes;Such That A'hy Additional SingleFailure Would Result in Station Blackout EVENT TYPE: Loss of Power -, '.*';-, , "

OPERATINGMODE APPLICABILITY: Run, Startup, Hot Shutdown ,

EAL THRESHOLD VALUE: ,.

The following conditions exist:

1. Unplanned loss .of power-o-,both Startup, ,(lX3) and ,Standby'.(lX4). transformers is expected to last for greater than 15 minutes.

AND. '

2.-Onsite&power capability has been degraded t6 one train 6f emergency busses powered from eitherA Diesel Generator (1G-31) or B Diesel Generator '(IG-21), and any additional single failurewill result in a St~tibh'Bladkout.-:

DAEC EAL INFORMATION: . - - " a - "

The DAEC EAL is .Jwritten to -address'the 'Underlyinrg" 'c6icernm, i.e., only o'nie :AC power source remains and if it is lost, a Station Blackout will occur. :'Under thte' conditions'6f concem, entry into AOP 301, Loss of Essential Electrical Power,' would be made under Tab ,1, Loss of One Essential 4160V Bus,, 6nd/oF'under -Tab 3;1 L6ss 6f Offsite Power.

Indications/alarms related to'degraded -AC,power are,,disolaliyed On control room panel 1C08 and are listed in AOP 301 under "Probable Indications." :" .. ..

At DAEC, 'the.Essential Buses of concern are 4160V Buses ,1A3 and 1A4. -Each of these" buses feed their associated 480V and 120V AC busses through step down transfoirners.

Onsite power sources at DAEC include the A and -B Diesel Generators, IG-31 'aid 1IG 21, respectively.

REFERENCES:

- - a - . *

1. Abnormal Operating Procedure (AOP) 301, Loss-of Essential Electrical Powerf -

2. UFSAR Chapter 8 Electrical Power . . - ".: '

3. Technical Specifications Section 3.8. Electrical Power Systems , , "

NEI Methodology for Development of EALs NUMARC/NESP-007 Revision 4, May 1999 SA5

V EBD-S SYSTEM MALFUNCTION CATEGORY- Rev. 3 Page 28 of 41 SSI Loss of.All Offsite Power and Loss of All'Onsite AC Power to

• Essential. Busses EP f r & - - - .

EVENT TYPE: Loss of Power OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown.

EAL THRESHOLD VALUE:

The following conditions exist:

1. Loss of power to both Startup (IX3) and Standby (1X4) ti'asformers.

AND -  ::  : "" " " ""

2. Failure of both A Diesel Generator (1G-31) AND B Diesel Generator (IG-21) to supply.

power to emergency bussees. I  ;- I -, I AND 1

3. 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.

DAEC EAL INFORMATION:

There is no significant, deviation from the generic EAL. In accordance with the generic guidance, DAEC -is using,.a- threshold, of 15 minutes for. Station. Blackout to exclude transient or momentary poWer losses.

Under the conditions of concern, entry into AOP 301.1, Station Blackout, would be made&

under Tab 1. Indications/alarms.related to station blackout'are displayed-on control'room panel 1C08 and are listed in the procedure, under "Probable, Indications."

REFERENCES:

1. Abnormal Operating Procedure (AOP) 301.1, Station Blackout.,

2., Technical Specifications Section 3.8, Electrical Power Systems

3. UFSAR Chapter 8, Electric, Power A "

4. NEI Methodology for Development of Emergency-Action Levels NUMARC/NESP-007 Revision 4, May 1999 SS1

EBD-S

-SYSTEM MALFUNCTION CATEGORYc,,.!'.' " Rev. 3 Page 29 of 41 SS2 .Failure of Reactor Protection System Instrumentatidn to.".,

"Complete'orInitiate'an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has .Ben Exceeded and Manual Scram Was NOT Successful EVENT TYPE: RPS Failure '-'- - -'-:

OPERATING MODE APPLICABILITY: Power Operation, Startup 4 .. 4 , -

EAL THRESHOLD VALUE:

Failure of automatic-scram and~actions6taken byoperators'in the CorntrolRoom to Shut..

down the reactorOR reduce reactor' power b~low the APRM dowhscales have been; INEFFECTIVE.

4 -ing " -* -xt l ~ h ~ _.'--' -'

The following conditions must exist to'debla' this EAL -,

1. In ATWS EOPI AND," -, '

2.- Operator actions to reduce power are UNSUCCESSFUL as indicated by either:

-a:Reactor power abovethe APRM Downsca/e Alarm on.ANY valid APRI .:

instrument, "

SOR.,,-, -.

b' Boron Injection Initiation Temiperature (Bl3lT)Curv, (FOP Graph6) exceeded.

DAEC EAL INFORMATION: . ' .

This -EAL addresses, c6nditi6ns ýwhere'failure ofan automatii scrdm tias'bccu6ried a'nd manual -actions -performed n',th ,:Control R6om to -reduhce reactor -,p0wer' have been unsuccessful.

Under the conditions of concern for this EAL, the reactor may be producing more heat than the maximum decay.heat load for which safety systems aredesigned. A Site Area Emergency is warranted because':conditions'exist that may lead to the potential loss of the fuel cladding or primary containment. Although this EAL may be viewed as redundant to the Fission Barrier Table, its inclusion is necessary to better assure timely recognition and emergency response.

SS2

The purpose of the ATWS EOP is.,to maintain: adequateý, core, cooling, shutdown :the, reactor and cooldown the RPV to- cold shutdown conditions. The ATWS EOP is implemented when it canrnot be determiihed thiat control'rod insertion alone will assure that the reactor will remain shutdown under all conditions.

Reactor power above the IAPRM"dow'nscale setpoint is indicative of power generation above the decay heat levels which primary containment is designed to suppress.

Furthermore, if reactor power is above the APRM downscale, setpoint, it is likely that the core bulk boiling boundary Would be above that which' provides suitable stability margin for operation at' high' powers and low flows.

Exceeding the Boron Injection Initiation Temperature (BIIT) limit (EOP Graph 6) is an indirect indication that the reactor, is at power and that excessive decay heat is being added to the suppression pool.

The higher the reactor power level is,the more heat energy will be rejected to the torus thus requiring a lower torus temperature for, ipitiation' of boron injection if the Heat Capacity Limit is not to be exceeded before reactor shutdown is achieved.'

As long as the core remains- submerged (the.preferred method of core cooling), fuel integrity and RP-V .integrity',*ie.e, .not directly challenged-, even: under failure-to-scram, conditions. However, a scram .failurejcoup!ed with. an, MSIV, isolation results. in rapid heatup of the'-torus dlu6e -to" the steam'discharged' from,the R1PV via SRVs. The challenge to the primary containment will thus become a limiting factor.

REFERENCES:

i . . ..

1. Integrated Plant Operating Instruction (IPOI) No. 5, Reactor Scram

2. ATWS Emergency Operating Procedure, (EOP) - RPV Control

3. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May1999 .. . , ,

SS2

EBD-S S , SYSTEM MALFUNCTION CATEGORY "'- -Rev. 3 Page 31 of 41 SS3- Loss'ofAll Vital DC Power - "-- -;

EVENT TYPE: Loss of Power .. 4-,.

OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EALTHRESHOLDVVAL'UE:' ' I 44 The following condition exists:,. .

1. Loss of both divisions of the Vital 250/125V. DC system based on AOP 302.1 and AOP, 388 for greater than 15 minutes.

.4. 4 -" , .. .

DAEC EAL INFORMATION: -' .  ;" ,. " " ;:,:.  :., -,

Under the conditions of concern, AOP 302.1, Loss of 125 VDC Power,.would be entered K under Tab' 3,; Complete Lbss of'125 VDC. Consequentlýi, the DAEC EAL-addresses' loss" of both divisions of the 125V DC'system consistent with AOP.

At DAEC, the 250V/125V DC Systems ensure 'power is available for-the reactor t'o be shutdown safely and maintained ina saf6'coriditiorin'..Th6 125V'S/ysterr'is'divided in'to tw6 independent 'divisions. -; Divisio "l ,andlDivision4- II - 'with "separate DC pidwEr 'upplies.

These power :supplies 6onsist'of1vt' 'sepratfe' 125'V'"b~tterieis -an'd "cll6rg*rs l6serMing systems such a§-RCIC; RHR, ED(s, and HPCI: -;"; -

Complete loss of both 125V DC Divisions could compromise the ability to-monitor and control the removal of decay heat during cold shutdown or refueling operations.

REFERENCES:

.. "4 ,."

Power _1

1. Abnormal Operating Procedure (AOP) 302:1, Loss of 125 VDC

2. Abnormal Operatinig Procedure'(AOP)'388;tEoss'of250VDC Po 'er'- 4 ' .4 -."

3. Technical Specification 3.8, Electrical Power Systems -,-

4. UFSAR Section 8.3, Onsite Power Systems

5. UFSAR Table 8.3-6, Plant Battery System ,- DC Power,- Instrumentation, ra'rd Control, Principle DC Loads (125V) -, " , - ,:,- '. °

6. NEI Methodology for Development of EmergencyAction Levels NUMARC/NESP-007 Revision 4, May 1999 - .  : -,

SS3

SS4 Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown EVENT TYPE: Inability to-Maintain Shutdown Conditions OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown.

EAL THRESHOLD VALUE:

1. EOP Graph 4 Heat Capacity Limit is exceeded.

OR

2. Reactor CANNOT be brought subcritical .

DAEC EAL INFORMATION: - .

This EAL addresses complete loss of ftunctons, including ultimate heat sink and reactivity, control, required for hot sh'utdoWn with, the reactor at pressure and. temperature. Under these conditions, there is an a*tual major failui'e of a system intended for protection of the public. The reactivity condition criteria is addressed- by maintenance of required, shutdown margin. If inadvertent criticality could not be eliminated by performing the actions of AOP 255.1, AOP 255.2, or the ATWSEOP, it corresponds to a failure of a system intended for the protection of the public and thus classification as a Site Area Emergency is warranted. ,

This EAL represents anr escalation from the conditions of concern in SA3, Inability to Maintain Cold Shutdown, because the reactor is at operating pressure and temperature and decay heat levels are higher.

Per DAEC Technical Specifications, the following systems are, necessary to achieve or maintain Hot Shutdown conditions:

"* Reactor Protection System Instrumentation

"* Core and Containment Cooling Systems Instrumentation

"* Reactivity Control

"* Standby Liquid Control System

"* Core and Containment Cooling Systems SS4

" Primary.System Boundary - '" '.

"* Auxiliary Electrical Systems  : - .. .. .,

Loss of instrumentation is addressed by SS6, Inability.to Monitora Significant Transient in Progress. The Auxiliary Electrical System is addressed by SS1, Station Blackout, and SS3, Loss of all Vital DC Power-and therefore they .aremqnot covered here. ýFailureof the.,

primary system boundary is covered by the Fission Barrier Table and SU5, RCS Leakage.

REFERENCES:

1. Abnormal Operating Procedure (AQP) 149, Lossof Decay Heat Remioval

2. Abnormal Operating Procedure (AOP),255.1,, Control Rod Movement/Indication 4 Abnormal ~ ~

3. Abnormal Operating Procedure (AOP) 255.2, Power/Reactivity Abnormal Change K-_> 4. Emergency Operating Procedure (EOP) 1 - RPV Control .

5. ATWS Emergency Operating Procedure (EOP) - RPV Control"

3. Emergency, Operating Procedure ALC

• AlternateLevel Control

7. Emergency Operating Proceddre ýEOO) Basi's,'EOP Brd'kpooinis 3: NEI Methodologi, fo5 Oevelopnfit bf Emerg6ncy Actioi Leels NUMARC/NESP.-070.

Revisior 4, May 19996 ':. .  :"' ..- . ' .- - . ,

. 'I a.. ~ .

SS4

SS5 Loss of Water Level in the Reactor.Vessel That Has or Will Uncover Fuel in the Reactor Vessel EVENT TYPE:- Inability to Mainfain' Shutdown Conditions OPERATING-MODE APPLICABILITY: 'Cold Shutdown, Refuel EAL THRESHOLD VALUE:

The following conditions exist

1. Loss of Reactor Vessel Water Level as indicated by:

a. Loss of all decay heat' " deterrnec'I.AOP 149 under Tab 1, Loss of Shutdown 6"hinigtn' AND I__

b. RPV level below 15 inches indicating that the core is or will be uncovered.

DAEC EAL INFORMATION: ,

The DAEC EAL is written in termns_.of the general concern.that no cooling water source is lined up or available for injection into the"RPV and'water: level is decreasing below the top of the active fuel (TAF). Under the. conditions of ccneem for EAL Threshold Value 1, AOP 149, Loss of Decay Heat Removal, would be entered under'Tabi'1, Loss of.Shutdown Cooling. Indications/alarms related to loss of shutdown cooling are displayed on control room panels 1C03 and 1C05 and are listed in the procedure. Consistent with the value used in the EOPs, the EAL uses an indicated RPV level of 15 inches for the water level corresponding to TAF.

The conditions address concerns raised by the NRC AEOD Report AEOD/EGO9, "BWR Operating Experience Involving Inadvertent Draining of the Reactor Vessel", dated August 8,1986. This report states:

In broadest terms, the dominant cause of inadvertent reactor vessel draining are related to the operational and design problems associated with the residual heat removal system when it is entering into or exiting from the shutdown cooling mode. During this transitional period water is SS5

SYSTEM MALFUNCTION CATEGORY ", - Rev. 3 4 .. Page 35 of 41

' drawn from the"reactor yessel; cooledby RHR heat exchangers° (fromf the cooling provided by the service water system), and returned to the reactor vessel. First there are piping and valves in the residual heat removal -,

system which are common to both the shutdowncooling mode and :other modes of~operation such as low pressure coolant injection and ."-- 'I"-

suppression pool cooling. These valves,when improperly positioned,ý provide a drain path for the reactor coolant to flow from the reactor vessel to the suppression pool or the radwaste system., Second, establishing or exiting the shutdown cooling mode of operation is entirely manual making such evolutions vulnerable to personnel and pr6ceduiral errors' Third, there is no comprehensive valve interlock arrangement for all the residual heat removal system valves that'could be activated during shutdown cooling. Collectively, these factors have contributed to the repetitive occurrences of the operational events rivolving theinadverteht.draining 'of the reactor vessel. ,..

REFERENCES:

. -,, . , , ., o..,  :. .6 . . ..

1. Abnormal Op[erating Procedure (AOP) ,1449, Loss of Decay Heat Removal- .

2. Emergency Operating Procedure (EOP)-1, RPV Control, Sheet 1 of1 "

3. Emergency Operating Procedure (EOP) Basis, EOP Break*poin"ts

4. NRC AEOD Report AEOD/EG09, "BWR Opratihg'Exerience lhvolving Inadveitent Draining bfth' Rea6torVessel",-Augist 8;'1986.. .

5." NEI Methodology f6i Develo6ment of Emergency Action LeV6els NUMARC/NESP-O07 Revision 4, May 1699'" i .

. * "L . - * " " '* ' " " ."' .. . ' ' . . . ". . " * ' " ' 4

.24

":, - ) .," ,'_:

• '. " -,  : .'"  ;' ,  : ;' -" *' L _,: .. _* *.; L -) .,)C

';.'444,  :' '"

SS5

SS6 Inabilityto Monitor a Significant Transientin'Progress EVENT TYPE: Instrumentation/Cdmmunication OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:

The following conditions exist:.

1. Significant transient in progress and ALL of the following:.

a. Loss of annunciators on Panels 1C03, 1C04 andil C05" AND - r'~V'

b. Compensatory non-alarming indications are unavailable.

AND

c. Indications needed to monitor criticality, OR. core heat ,removal, OR Fission Product . '

Barrier status'are unavailable.

DAEC EAL INFORMATION:

The DAEC EAL is written in terms of a significant transient in progress with loss of both safety system annunciators and loss of compensatory non-alarming instrumentation. The DAEC EAL structure, which addresses all the key points in the generic EAL, better assures that the condition of concern for this EAL will be readily recognized.

Significant transient includes response to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or undamped thermal power oscillations greater than 10%.

Compensatorynon-alarmingindications include the plant process computer, SPDS, plant recorders, or plant instrument displays in the control room. These indications are needed to monitor (site-specific) safety functions that are of concern in the generic EAL.

SS6

EBD-S

-.,FSYSTEM MALFUNCTION CATEGORY ' .Rev. 3 S .- Page 37 of41 Control room panels , C03 .-l C04,- and 1Co5.conhtini the annunciators associat6d with safety systems at DAEC. Annunciators on 1C03, 1C04, and lC05'share .a common power supply from 125 VDC Division I that is -,fed. through circuit breaker 1D13.

Therefore, DAEC does not specify a loss of "most" annunciators as specified in the generic methodology. . .

Indications of loss of annunciators associated with safety systems include:

• 125 VDC charger, battery, or system annunciators on control room pan'el 1*C08 0 Loss of "sealed in" annunciators at affected panels , .

• Failure of affected annunciator panels shiftily testing by plant operators 0 Expected alarms are not received , -. ,,*t -, - - -.

• 0 Computer point ID B350 indicates "NSS ANN DC LOSS TRBL." (Loss of DC.power to panels 1C03,1CO4, and CO5)

REFERENCES:

1. Operating Instruction (01)VNo. 317.-2,,Annunciator.System

2. Abnormal Operating Procedure (AOP) 302.1, Loss of 125 VDC Power

3. Abnormal Operating Procedure (AOP) 302.2, Loss of Alarm Panel Power

4. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May 1999

-:., II ,, * -. ,- , ,,I.

SS6

I I

-S SYSTEM MALFUNCTION CATEGORY; Rev. 3 Page 38 of 41 SGI Prolonged Loss of All Offsite Power and Prolonged Loss of All Onsite AC Power - ':

EVENT TYPE: Loss'of '

Power

• f* ' * " I OPERATING MODE APPLICABILITY: Run, Startup, Hot Shutdown EAL THRESHOLD VALUE:

The following conditions exist:

1. Loss of voltage on buses 1A3 and 1A4.

AND ANY ONE OF THE FOLLWING

a. Restoration of power to either Bus 1A3 or 1A4 is' not likely within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. N OR .

b. RPV level is indeterminate. - I OR " '

c. RPV level is below +15 inches.

DAEC EAL INFORMATION: "

There is'no significant de6i6i`6n frm the gdrieric EAL. Under prolonged Station Blackout (SBO) conditions, fission':p~oduct pbarrie* 'monit6rng *capabilitý may be 'degraded.

Although it may be difficult to predict when power can be restored, it is necessary to, give the EC/OSM a reasonable idea of how quickly a General Emergency should be declared based on the following considerations:

• Are there any present indications that core cooling'is already degraded to the point where a General Emergency, is IMMINENT (i.e., loss of two barriers and a potential loss of the third barrier)?

If there are presently no indications of degraded core cooling, how likely is it that power can be restored prior to occurrence of a General Emergency?

SG1

The first part of this EAL corresponds to the threshold conditions for Initiating Condition SSI'-Station Blackout -. namely, entry into AOP 301.1, Station Blackout. The second part of the EAL addresses the conditions that will e&scalate the SBO'to General'f-Emergency.

Occurrence of any of the following is sufficient for escalation: (1) SBO copinig capability exceeded, or (2) loss of drywell cooling.-that continues to ,make RPRVwater level measurements unreliable, or (3) indications of inadequate core cooling.' Each of these conditions is discussed below: .... ,

1. SBO Coping Capability Exceeded '

DAEC has a SBO coping duration of four hours.' The likelihood of restoring at least one emergency bus should be based on a reafisticdappraisal6f the 'situati6n 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 preparingand implementing publicprotective actions. -.

K> 2. RPV Water Level Measurements Remaining Unreliable Flashing of the reference leg water will result in erroneously high RPV water level readings giving -a false indication of actual water, inventory and potentially indicating adequate core cooling when it may not exist. EOP Graph 1, RPV Satuiration Temperature, defines the conditions under which RPV level instrument leg boiling may occur. .. - - - '

3. Indications of Inadequate Core Cooling DAEC uses the RPV level that is used for the.Fuel Clad "polential los$" condition in the Fissidn Product Ba'frie'r Mahtrix.- This is RPV level below +1.5 inc'es.

REFERENCES ,. . .

.... .. .nc..

. Abnormal Operating Procedure (AOP) 301.1, Station Blackout - . ,

2. .Letter NG-92-0283, John F.,Franz, Jr..to Dr.-Thomas E..Murley, Responseto Safety Evaliuation by*,NRC-NRR -Statiodn' Bl6c~ko'utE"valuation Iowa" ElectricLight and Power Company Duane'Arold Eneiby Center," February , 1992, 7 - ,] -.

3. EEmergency Operating Procedure (EOP)I - RPV Control -. - ,.

4. Emergency Operating Procedure,(EOP)ALC -Altemate Level Control 5.- NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007,,

Revision 4, May19 199.9 -. , , , - - , -

L SG1

SG2 Failure of the Reactor Protection System to Complete an, Automatic Scram and Manual-Scram was NOT Su6ccessful and' There is Indicationof an Extreme Challenge to the Ability to Cool the Core - . '"

EVENT,TYPE: RPS Failure

• OPERATING MODE APPLICABILITY: Power Operation, Startup EAL THRESHOLD VALUE: '

Failure of automatic and manual scrams AND conditions exist" thao 16hg6ý assure adequate core cooling or adequate decay heat removal.

The following conditions must exist to declare this EAL:

1. InATWSEOP AND',

2. Loss - '-

b"6-ade-d.,e.,o. d;cayhe'at'r ,

2. ng heat removal capabilityi as indicated by eitlýe-,

a. RPV level cannot be restored and maintained above the Minimum Steam Cooling RPV Water Level (i.e., SAG Entry Required),

OR -

b. HCL Curve (EODPFGraph ,4) exceeded. -.

DAEC EAL INFORMATION:

This EAL addresses conditions where failure of an automatic scram has occurred and' manual actions performed in the Control Room to reduce reactor power have been, unsuccessful AND a subsecuent loss of adequate core cooling o4 deday heat removal capability occurs. -If either of these challenges exist -during an ,ATWS, a core melt seq'ulence exists. In this situatiori 'core degradation can occur rapidly. For this reason, the General Emergency declaration is' intended to be anticipatory of the fission product barrier matrix declaration to permit maximum offsite interventi6n time.

The purpose of the ATWS- EOP is to maintain adequate core cooling, shutdown the reactor and cooldown the RPV to cold shutdown conditions. The ATWS EOP is SG2

'* '-**!  ;*EBD-S

-`:SYSTEM MALFUNCTION CATEGORY*  :*'. Rev. 3 Page 41 of 41 implemented when it cannot be determined that control rod'insertion alone will assure that the reactor will remain shutdown under all conditions.'.

If injection whall -available 'Preferred and Alternate ATWS Injection Systems fails to provide sufficient injection to restore and maintain level above -25 inches (Minimuim' Steam Cooling RPV Water Level), adequate core cooling is threatened and.submergence of the core is attempted by flooding the primary containment. This is accompl ished by transfer to and implementation of the DAEC Severe Accident Guidelines (SAGs).

The Heat Capacity Limit (EOP1 Graph 4) is defined to be the highest torus temperature at which initiation of RPV depressurization will not result in -exceeding the Primary Containment Pressure Limit (the PCPL is 53 psig at the DAEC) before the rate of energy transfer from the RPV to-theiprimary containment is within the capacity of the containment vent.' - - , ,.* - .. , .

Control of torus temperature relative to the Heat Capacity Limit is directed in the Primary kj*) Containment Control Guideline, EOP 2. If the actions being taken in EOP 2 to preserve torus heat capacity are inadequate or not effective, RPV pressure must be reduced in order to remain below the Heat Capacity Limit. Therefore, actions in the RPV pressure control section of the ATWS EOP must accommodate these requirements. Failure to do so may leaid to failure 'of the 'containment or loss of 'equipment necessary for the safe shutdown of the plant.

REFERENCES:

1. Emergency Operating Procedure ATWS EOP - RPV Control

2. NEI Methodology for Development of Emergency Action Levels NUMARC/NESP-007 Revision 4, May1999 SG2