Text

Rev 0 to Emergency Action Levels Technical Basis Document. W/Two Oversize Encls
	ML20046A478
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Site:	Calvert Cliffs
Issue date:	06/15/1993
From:	; BALTIMORE GAS & ELECTRIC CO.
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B ALTIMORE ! GAS IND E L E C T R I C T- : CALVERT CUFFS NUCEAR POWER PLANT UNITS 1 & 2 i

EMERGENCY ACTION LEVELS TECHNICAL BASIS DOCUMENT REVISION O June 15.1993 f PEPAND BW g

                                                                   -        DA*E:       !      3 Ogoen Prqy Manager E1EWED BY:           fy                      g(                      DAE   h  '
                                                                                          /     3 Emergency Planning Gmergency Pianning Anatyst NVEWED BY:

DAE $ 83 73 Planti Oper art Supe iso L haconsfd$r~r - dovb : toececons Requaiticacon Trainfg - Instructor . AEMEWED BY: / DAE /p - -M Chemistry Programs- Chemist M Bt y gy y e_f _- DAE & .2 y.O J Macacon Safety- Sr./ Plant Heartn Physicist AEVEWED BY: DAE 6 [J 7 /9J NucTear Engineenng ' Engineer WM Bt f,9 3 ,jf , y y j y ,,,, ,J & DAE fh ' Q3 Secunty - Secunty Programs, Specialist nc-o st 92 h DJi 3 Nuclear egulatory Anaryst DAE c h2/e

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APPA0VED Bt ny[t. j DAE h.Ed.45 Emergency Plamning Unit - Supervisor TVIEWED Bt- I [ p j3-j'-f DAE 74-{J POSRC i 2 jr / Meeung No. Y

APPACNED Bt ,e/ g g/t, g;D -

DATE: /1 N5

  % ,/              PtareGeneral Manager

TABLE OF CONUS GENERAL NOTES FOR EAL TECHNICAL BASIS . . . . . . ... . .. . . ... G;1 RADIDACTMTY RELEASE O- AU1 Unplanned Radiosaue Release Exceeding 2 X Tech Spec Umits for AT LEAST BO Minutes . . R1 RU2 Unexpected incmase in Rant Radiation . ..... .... ... . ... .. . ., ... R.5 RU3 Potential Degradation of Containment of Dry Stomd Spent Fue! .. ... .. .. . R:7 RA1 Unplanned Radioactive Release Exceeding 200 X Tech Spec Umits for AT LEAST 15 Minutes . . . R:9 RA2 Damage OR Uncovery of Single Irradiated Fuel Assembly Outside the Reactor Vessel . . .. R:13 RA3 Radiation increases That impede Safe Rant Operation . . .. ... .... ..... . R:15 RS1 Off-Srte Dose of AT LEAM O.1 Rem (EDE + CEDE) Or 0.5 Rem CDE Thymid . .. R:18 RG1 Off-Site Dose of AT LEAST 1 Rem (EDE + CEDE) Or 5 Rem COE Thyroid .... . . . . . R22 RSSION PRODUCT BARRIER DEGRADATION BU1 Loss OR Potential Loss of CNTMT Barrier . . .. . 8:1 BU2 RCS Leakage . . . ..... .. . ... B2 BU3 Fuel Ctad Degradation . . . . B:4 BA1 Loss OR Potenual Loss of EITHER Fuel Clad Bamer OR RCS Barvier . B.B BS1 Loss Or Potential Loss of ANY Two Bamers . ... . B:7 BG1 Loss of Two Baniers AND Potential Loss of Third Barrier . . . .. B:10 FUEL CLAD BARRER EALs .. .. . . . . B:11 FCB1 Safety Function Status / Functional Recovery . . . . B:12 FCB2 Temperatum . . . .. B:14 FCB3 Radiation . . . . . . . . . . B:15 FCB4 Reamor Vessel Water Lsvel . . . . .. . . . .... 8:17 FCBS SEC Judgement , .. . B:18 RCS BARRER EALs .. . . .. . . . . . B:19 RCB1 Safety Function Status /Funaional Recovery . . .. .. B-20 RCB2 Temperature . . . .. .. B22 \ RCB3 Radiation . . . B24 RCB4 Coolant Leakage . . .. . ... . . . .. 825 RCBS SEC Judgement . . .. . . .. . . B27 CONTAINMENT BARRER EALs . .. . . . B28 CNB1 Safety Function Status /Funcdonal Recovery .. . . 829 CNB2 Temperature . . .. .. . .. .. .. B:31 CNB3 Radiation . .. . . 8:33 CNB4 Coolant Leakage . .. . . . B35 CNB5 Pmssure ... . . . . . .. . B:37 CNBB SEC Judgement .. . . .. . . B:39 EQUIPMENT FAILURE QU1 Unplanned Loss of Any Function Needed to Maintain Cold Shutdown Q.1 QU2 Unplanned Loss of Most or All Safety System Annunciators for GREATER THAN 15 Minutes . Q3 QU3 Unplanned loss of All On-Sce or Off-Sce Communications Capabilcies . . 25 ; QU4 Inabiley to Reach Required MODE WEhin Technical Specifcation Umits . . .. . G7 j GA1 Failure of Automatic Reactor Trip . .. .. . .. . . . . Q.B i QA2 inabilty to Maintain Plant in Cold Shutdown . . . . .. . . . Q10 1 QA3 Unplanned Loss of Safety System Annunciators Wch Transient in Progress . Q12 ' QS1 Failure of BOTH Automatic AND Manual Reactor Trip . . Q14 ! QS2 Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown . CL15 QS3 Loss of Water Level That Can Uncover Fuel in the Reactor Vessel . .. O17 QG1 Failure of BOTH Automatic AND Manual Reactor Trip -AND-Extmme Challenge to the Abilty to Cool the Core . . Q19 l Cakert Ctrffs EAL Basis Document i June 15,1993

TABLE OF CONTENTS l EU2 Loss of Veal 1'25 Volt DC Power for GREATER THAN 15 Minutes . . . . . . . . . . . . . . . . . . . . . . . . . E3 l EA1 Station Blacko2 While On Shutdown Cooling . . . . . . ..... . . .... ...... . . . . . . . . E7 ' EA2 Ony One AC Power Source Available to Supply 4 kV Emergency Buses . .. ... ..... . . . EB EA3 Loss of 123 Vot DC Power AND Reactor Trip . . . ... .... .. ........ . .... E:10 (q) ES, Steeon B w out.... . ........... ...... .... . ......... ...... ... ..... e,, ES2 Loss of All 125 Volt DC Buses . . . ................ ....... . ... . ... . .. E12 ; E14 EG1 Pmionged Station Blackout . ... ........ ...... .. ...... ..... . ... .... SECURITY ............ ...... .. ... ......... .... ....... ...................To TU1 Confirmed Security Event With Potential Degradaten in the Lsvel of Safety of the Plant . . ... T;1 : TA1 Security Event in the Plant Pmtected Ama . . . .....................................T2 TS1 Security Event in a Plant Vital Area . . . . . .. .. .......... .. . .. ... .. . T;4 TG1 Security Event Resuting in Loss of Abilty to Reach AND Maintain Cold Shutdown . . . ... ..T:5 RRE.. . .. . . . . . .. .. .. . . . . .. . 10 IU1 Rre Wsthin Pmtected Ama Boundary Not Extinguished Wohin 15 Minutes of Detection . . . .. L1 1A1 Rm or Explosion Affecting Safe Shutdown .. . . . . . .. . .. . .. .. .. L3 NATURAL PHENOMENA . . . . . . .. . .. . . ........ . .. .... ... NO NU1 Natural Phenomena . . . . . . . . . . . . . . . . . . . .. .. . . . N:1 NA1 Natural Phenomena Affecting Safe Shutdown . . . . N:4 OTHER HAZARDS ... . ... . .. ... . .. . .. . ... . . OD OU1 SEC Judgement .. .. . . .. .. .. . 0:1 OU2 Toxic or Flammable Gases . . .. .. .. . ..... . .. ..... .. .. . .... . 03 OU3 Destructive Phenomena . . . .. . . . .... . .. . . .. . . 0:4 OA1 SEC Judgement . . . . .... . . . ... .. . .. .. . 0:7 OA2 Toxic or Rammable Gases Affecting Safe Shutdown . .. .. . ... ... . 0.8

 '  OA3 Destructive Phenomena Affecting Safe Shutdown                          .         . . ... . .... ....... . . ..                                          0:10 OA4 Contml Room Being Evacuated . . . .                       .                . .....               . .. . ....                       . . . . . . 0:12 OS1 SEC Judgement .              .                        .       . . .                                ..              .           .          .             O.13 OS2 Contml Room Has Been Evacuated AND Timely Plant Control Can NOT Be Established                                                                . . . D:14 OG1 SEC Judgement          .         .     .                  .            ...          .       ..              .           .. .. . ..                  . 0-16 i

i s Cah/ert Cliffs EAL Basis Document ii June 15.1993

TABLE OF CONTENTS TABES n Table G-1: Comparison of NUMARC Guidelines to BG&E Cs fJJMARC Atrormal Rademon LeeWRadiological Effluert Category . . . . . . . . . . . . G:6 (% Table G2: Comparison of NUMARC Guidelines to BG&E Cs PUAARC Hazarts and Other Condoorm Af'ecung Piert Safety Category . .. . . G.7 Table GG: Comparison of NUMARC Guidelines to BG&E Cs PDAAAC System Malfuncuan Category . . . . . . . . . G:B Table G4: Comparison of NUMARC Guidelines to BGSE Cs taJuARC Rasion nuduct Bamor co;rademon Category . . . .. . . . G:10 Table G5: Comparison of NUMARC Guidelines to BG&E EALs PUAAAC Ar. son Product Bamor Degradauon Category . . . .. . .. .. .. . G:11 Table B-1: SAE Bamer Loss / Potential loss Combinations for CCNPP Logic . . .. . B:8 Table BG: SAE Bamer Loss / Potential Loss Combinations for NUMARC Logic , . ... . 8:9 Table E-1: Effects of Lost 125 Volt DC Buses 11,21,12, and 22 .E:5 C ( Cabert Cirtfs EAL Basis Document iii June 15.1993

i O m GENERAL NOTES FOR EAL TECHNICAL BASIS O t

GEtEJML NOTES FOR EAL TECHNCAL BASIS The following general notes apply to the Calvert Chffs EAL Technical Basis information: q 1. The fortnat of the EAL Technical Basis information was developed to address training needs, to facittate NRC approval, and to facilcate future revisions and 10 CFR 50.54(q) evaluations.

2. NUMARC generic informatiors is quoted directly fmm NUMARC/NESP OO7, Revision 2. dated January 1992.

Changes from the NUMARC taxt are denoted by carut marts (< >). Such changes are based on the following enteria:

To put the NUMARC generic information in its pmper context such as when k; refers to a section of the NUMARC document.
To rename inciating Condcions (Cs) fmm their NUMARC designation to the corresponding Calvert Cirffs designation.
To delete information that does not appFy such as reference to BWR information.

u

3. The EAL Technical Basis information is organized by Event Category which is shown by the Tcle on each page:

_Radioactkey Releases (R)
Fission Pmduct Barrier Degradation (B)
SecunTy (T)
Equipment Failure (G)
Fjre (1)

_ Naturel Phenomena (N)

Electrical (E)
_Other Hazards (0)

Calvert Cirffs designations use two letters followed by one number. The identifier numbers wem selected so that they would not overlap wth NUMARC C designators and thereby cause confusion. The forst letter corresponds to the event category as shown above. The second let%er corresponds to the emergency classifcation level for the C:

U - (Notifcation of) Unusual Event A-[Uert
S - S.te Emergency
G - General Emergency The number designates whether the C is the fir % second, third, etc., C for that event category under that emergency classification. For example, BU2 is the identifier number for the second Fission Product _ Barrier Degradation category C in the Unusual Event classifcation. EG1 is the first FJectncal category C in the General Emergency classifcation, etc.

Similarty. Cabert Cirffs Fission Barrier EALs also use dfferent designators than NUMARC. They are: FCB _ Fuel Clad _ Barrier RCB - RCS _Barner CNB - con:ainment _Barner

4. Catvert Chffs Operational Modes are referred to by the cortesponding Technical Specrfcation Table 1.1 numbers. These are:

Mode 1 - Power Operation
Mode 4 - Hot Shutdown
Mode 2 - Startup
Mode 5 - Cold Shutdown
Mode 3 - Hot Standby
Mode 6 - Refueling Calvert Chffs EAL Basis Document G.1 June 15,1993

GENERAL NOTES FOR EAL TECHNCAL BASIS Please refer to the Tech Spec table for the corresponding temperatum, pressure, and reactivty parameters for each of these operational modes. Operational modes applying to Cs/EALs are based on the opemtional mode that the plant was in immediately before the event sequence leading to entry into the emergency classdication. For example, events /condtions addressed by Cs appleable to Mode 1 (Power Operation) are expected to lead to reactor trip which should bring the plant to Mode O (Hot Standby). However, the appmpriate emergency classifcation would still be based on the appleable Cs for Mode 1 operation for these events / conditions.

5. The " Plant-Specife Basis

section for each C/EAL pru/ ides the description of how NUMARC genene information was applied to develop Cakert Cittfs EALs. Supporting pmcedures, calculations, their undertying bases and assumptions. and their results are fully described in the "Plantepecife Basis
section, as appmpnate.

6. Fmquently used terms are defined below-AC- Alternating Current i

Alert - Events are in process or have occurved which involve an actual or potential substantial degra dation of the level of safety of the plant. Any releasIs are expected to be limited to small fractions of the EPA ; Protective Action Guideline exposure levels. All- Applies to Operational Modes 1 thmugh 6 (listed above) plus defueled mode. l AT LEAST- Greater than or equal to ATWS- Anticipated Transient Wohout Scram , Barrier- Same as Fission Pmduct Barrier below. Barrier Monitanng Abilrty- This must be considered as an SEC judgement factor in determining whether a fission product barrier is lost or potentially lost. Decreased abilty to monitor a banier results from a loss of/ lack of miiable indcators, including instrumentation operrbilty concems, madings fmm portable instrumentation, and consideration for offste monitoring results. Can NOT- The final safety function status is of concem, not merely the inabilty to meet certain intermediate status check conditions. CT- Committed Dose Equkalent as defined in 10 CFR 20.1003 CEA - Contml Element Assembly

      &M- Committed Effective Dose Equivalent as defined in 10 CFR 20.1003 GDM - Contml Element Drive Mechanism ET- Core Ext Thermocouple CFM- Cubic Feet per Minute CNTMT- Containment Comperssatory normlarming indications - includes computer based informatson such as SPDS. This should include all computer systems available for this use depend!ng on spectfc plant design and subsequent retmfts.

Cabert Citts EAL Basis Document G.2 June 15,1993

GENERAL NOTES FOR EAL TECHNICAL BASIS CPM- Counts Per Minute

-     CSFST - Critical Safety Function Status Tree CST- Condensate Storage Tank DAC- Derived Air Concentration DC- Direct Current Dominant accident sequences - These will lead to degradation of all fission pmduct barriers. They include ATWS and Station Blackout sequences that am separately addressed under the Equipment Failure and Electncal categories, respectively, as well as by the Fission Product Barner Degradation EAL Tables.

ECCS- Emergency Core Cooling System EUE- Effective Dose Equivalent as defined in 10 CFR 20.1003 Emergency Action Levels (EAU - A pre <ieterTnined, site-specific, observable threshold for a plant initiating Condition that places the plant in a given Emergency Class. An EAL can be; an instrument mading, an ; equipment status indicator, a measurable parameter (on site or off-site), a disemte observable event, results ! of analyses, entry into specific emergency operating procedures, or another phenomenon which, if it occurs, ! indicates entry into a particular Emergency Class. I f Emergency Class - Same as Emergency Classifcation Level below. l Emergency Classification Level-These are taken fmm 10 CFR 50 Appendix E. They are,in escalating order. (Notification of) Unusual Event (UE), Alert, Site Emergency, and General Emergency (GE). Assion Product Bamer- One of the three principal barriers to uncontrolled release of radionuclides: Fuel Clad, Reactor Coolant System (RCS), and the Containment building (CNTMT) ; FDST- Fuel Oil Storage Tank ; t Fuel CIad- The zirconium alloy tubes that contain the fuel pellets. ! GeneralEmergency(gel- Events are in process or have occurmd which involve actual or imminent substantial com degradation or melting with potential for loss of containment integrity. Releases can reasonabty be , expected to exceed EPA Protective Action Guide (PAG) exposure levels off-site for more than the immediate l site ama. GPM- Gallons Per Minute ; inadvertent - Accidental or unintentional, e.g., the event occurred because procedures were not strictly adhered to. Amminent - Refers to anticipated degradation of any fission pmduct barrier within 2 hours based on a projection of current safety system performance. lo service - A component or system in the appropriate configuration for norinal operation and is considered

     " operable" as defined in the Carvert Cliffs Technical Specifcations Section 1.B.

Initiating Condition (ICl- One of a predetermined subset of nuclear power plant condcions where either the potential exists for a radiological emergency or such an emergency has occuited. l KV (kV)- Kilovolts, i.e., thousand volts j i i Catvert Cirffs EAL Basis Document G:3 June 15,1993

GEERAL NOTES FOR EAL TECHNICAL BASS LOCA - Loss of Cnolant Accident loss (of a fission product bemer)- A severe challenge to a fission pmduct barrier exists such that the barrier is considered incapable of performing lts safety funa. ion. Millirem - One thousandth of a mm MPH- Miles Per Hour NOT Effective - Corrective ad. ions do not yield appropnate or satisfactory msults based on available operable instrumentation madings. Noufication of Unusual Event (NOW]- Same as Unusual Event below. Planned- Loss of a component or system due to expected events such as scheduled maintenance and testing acuvrties. Potential Loss (of a fission product bamer)- A challenge to a fission pmduct barrier exists such that the barrier is considemd degraded in its abiley to perform its safety function. PSIG- Pounds per Squam inch Gauge PTS- Pressunzed Thermal Shock PWST- Pretreated Water Storage Tank PZR - Pressuruer RCS- Reactor Coolant System RCP- Reactor Coolant Pump Ram - Unit of radetion dose as defined in 10 CFR 20.1004 Requimd- Entry into a given procedum is neither optional nor memy suggested; rather, it is imperative based on existing condtions. RFP- Refueling Pool RTP- Rated TherTnal Power RVLMS- Reactor Vessel Level Montoring System RWT- Refueling Water Tank l SDC- Shutdown Cooling ; I SDCS- Shutdown Cooling System i SEC- See Emergency Coordinator SG- Steam Generator ; l Sievert (Sv]- Unit of radiation dose equivalent to 100 mm j O ! l l l i Cakert Cliffs EAL Basis Document G4 June 15,1993 l l

GEfERAL NOTES FOR EAL TECHNICAL BASIS Signilicant emnsient - (See also, " Transient", below.) includes msponse to cutomatic or manually initiated functions such as scrams, runbacks involving gmater than 25% thermal power change, ECCS injections, or thermal power osci!!ations of 10% or gmater. L SfT - Safety injection Tank Site Area Emergency (SAEJ - Events are in process or have occurred which involve actual or likely major failures of plant functions needed for protection of the public. Any releases are not expected to result in exposure levels which exceed EPA PmtectNe Action Guide (PAG) exposure levels except near the site boundary. Site Emergency- Same as Site Area Emergency above. TEM- Total Effective Dose Equivalent as defined in 10 CFR 2D.1003 Transient - A condltion that is:

Beyond the expected steadystate fluctuations in temperature, pressure, power level, or water level, and
Beyond the normal manipulations of the Control Room operating crew, and
Expected to require actuaton of fast-acting automatic control or protection systems to bring the reactor to a new safe, steady-state condtion.

Uncontmlied means that gNen condition is not the msult of planned actions by the plant staff. Unisolable means that actioris taken from the Main Control Board or locally are not successful in eliminating the leakage path. Unplanned is used to pmelude the declaration of an emergency where a component or system has been removed intentionally fmm service (e.g., for maintenance and/or testing activites). As used in the context of rad releases, ' unplanned" includes any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions (eg., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. Unusual Event (UEJ- Events em in process or have occurmd which indicate a potential degradation of the level of safety of the plant. No releases of radioactive material requiring off-site response or monitoring am expected unless further degradation of safety systems occurs. Valid means that the indication is from instrumentation determined to be operable in accordance with the Technical Specifcations or has been venfied by other independent methods such as ind'cations displayed on the control panels, reports from plant personnel, or radiological survey results. WRNGM- Wide Range Noble Gas Monitor k Calvert Cliffs EAL Basis Document G.5 June 15,1993

GENERAL NUTES FOR EAL TECHNCAL BASIS Table G 1: Comparison of NUMARC Guidaines to BG&E Os NUMARC AbncxTnal Redistion twels/Rediological Effluent. Category Emergency Osss Generic NUMARC C Calvert Diffs C Unusual Event AU1 - Any Unplanned Rdosse d Gaseous or RU1 - Unplarrod Radioac2wo Rolosse Exceedng l Uguid Radeoectuty to the Erwunmort That 2 X Tech Spec Umes for AT LEAST 60 Mnutes Ex: cods Two Trnos the Radiologk:al Techrscal Spuv% = for 60 Mnutes or Longer l AU2 - Uncxpected increase in Piart Radaborro RU2 - Unexpected increase n Piart Radetaan RU3 - Potarcel Degredation of Cortarrnort of ; DY Stored Sport Fuel l Alert AA1 - Any unplanned Rolesse of rem or RA1. Urplanned Radioac2Ne Release Exceeding i bai.xd RadioactMty to the Erwunmort That 2CD X Tech Spoc Umts for AT LIAST 15 Exmods 2CD Tmos Radolagical Technical M outes Specrfications for 15 Moutes or Lunger AA2 - Mapor Damage to kredated Fuel or loss RA2 - Demsge OR thcrwory & Sngle Irmdated of Water level That Has or Will Result n tre Fuel Assomtdy Outado tre Reactor Vessol Urocwonng of hudaad Fuel GAude the i Reactor Vessel AA3 - Roease of Rodoacthe Mstenal or RA3 - Redenon anarnasos That impode Safe l bcreases in Redisuon Lemis Wchn the Fecky Plert Operstaan l That impodos Opvstion of Systems Roqured to Martan Safe Operatons or to Estabish or I Maintain Coid Shutdown ! I Site Emergency AS1 - Boundary Dose Resubng From an Actual RSi - OffGte Dose of AT ifAST Q1 Rem [EDE or invrsrort Reicase of Gassous RadioactMty + EDE) & G5 Rom CDE Thrud i Excoods 1CD mR wtioie Body or SCD mA Ould l Thrud for the Ac2ual or Prugeted Durnoon of I f the Reicase l

 \    General Emergency       AG1 - Boundary Dose Resubng From an Actual    RG1 - Off&te Dose of AT LEAST 1 Rom [EDE or Inynnort Rolosse of Gaseous Re60ectuty     +2DE) & 5 Rom (CDE) Thrud Exceeds 1000 rrA Whoie Body cr 5000 rrA l

Chad Thrud for the Actual or Pruinc:ad CLrsuon of tre Re6 ease l 1 l l l l l i Calvert Cliffs EAL Basis Document G6 June 15,1993

GENERAL NOTES FOR EAL TECHNCAL BASIS

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Table G2: Compenson of NUMARC Guidelines to BG&E Cs NUMARC Hazards and Other Conditions Affeedng Plant Safety Category Emergency Dess Generic NUMARC C CeNert Drffs C Unusual Event 101 - %: ural and Dostnxsave Phemmene OU3 - Destructm Phenornene Affectng the Pmtocted Arte MJi - Natural Phenomore FU2 Frs Withn Protected Area BW Not U1 - Fire Wthin Prutacted Arve Boundary Ntt Exbngushed withn 15 Mnutas of Dotatxn Ecngushed Wchn 15 Minutna of Detectaan FU3 - Ro6sase d Tome or Rarrmede Geses DU2 -Tome or Rammablo Gasos Deemod Datnmortal to Safe Operstaan d the PLart FU4 - Conferned Secunty Ewrc Whch hdcatas TU1 - Confrmed Socmty Evert With Potortaal e Potanual Cow udou, n the Lawnl of Safety of Dogradataan in the Lawi of Sefoty of the Ptert the Ptart HU5 - Other Condcors Emstang Whit e/W Oui - SEC Judgemort Judgomo t of ste Emorgency Dructor Warrst Declaeon of an unusual Ewrt Alert, FM1 - Natur31 and Dostnxxave Phenomore OA3 - Destructm Phenomons Affectang Safe Nfecong the Plert Veel Arse Shutdown tai - Naturst Phonomone A#cctang Safe Shutchn HA2 - Are or Explosion Af+ectng the Operstacy M1 - Are or Expi:xuon Affectang Safe Shutzlown d Ptert Sefoty Systems Requred to Estatds5 or Maintan Sete Shutzhn FM3 - Release of Tocc y Rammebe Gasos OA2 - Tome or Rarrmatdo Gases Affectang Safe Withn e Fecsley Stnxnri Whch Joopenizes Shutdown Operwuon d Systems Requred to EstatAsh or Mairtain Codd Shutdown FM4 - Socxanty Evert in a Plat Pmtectad Area TA1 - Secunty Evert in the PLart Protected Area HA5 - Cored Room Eweaus:aon Has Boon OA4 - Cored Room Bong Ewcaastad Ircated FMS - Other Conderans Emsung mch n the DA1 - EC Judgomort Judgomort of the Emorgency Dructor Wenwt Doctemon of an Alert I Srte Emergency HS1 - Secunty Event n Pnart Vtal Area TS1 - Secunty Evert in a Ptert Vtal Area l HS2 Coed Room Evacueuon Has Boon W2 - Cored Room Has Boon Evecustad and l irsusted and Ptart Contrd Cannot be EstatAshed Trrely Port Contrd Con FUT Be EstatAshed l l FG3 - Other Condtaans Emstang Whe in tre OS1 - SEC Judgomort Judgoment d the Emorporcy Dructor Warrwt Dodseon of a See &= wo m General Emergency FG1 - Secunty Evurt Resubng n Loss of Atmicy TG1 - Socunty Evert Resubng n Loss of Atmicy to Roach and Martain Co4d Shutzhn to Roach 40 Marten Cold Shutdown FG2 - Other Condoorm Emstang Whd in tre D31 - SEC Judgomort Judgomort d tre Emerporry Drector Womrt Doctaeon of a General Emerporcy I Cakert Cir!fs EAL Basis Document G:7 June 15,1993

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(BERAL hDTES FOR EAL TECHNICAL BAS:S . - - f Table G3: Compenson af NUMARC Gudelines to BGGE ICs , NUMARC System Malfurx: tion Category Emergency Qass Genanc NUMARC C CaNort Qiffs C i I Unusual Ewnt Sui - Loss of M Offeite Pomer to Essentaal EU1 - Loss of Offste Power Busses fx Greater Then 15 unutes & SU2 -Inabihty to Reed Requeid Shutdoun GU4 - habihty to Reach Requned MDCE Within , Wthn Tectnical ^- '..- - , Umta Techncel R- E a- .Urnits SU3 - Urpismed Loss of M Safety Sptern GU2

Unplemed Laos of Most w M Safety Annurostors fr Groeter Then 15 Mrutas System Anntmetars for GREATER THAN 15 Mnutes SU4 - Fuel Osd Dw .h BU3 Fuel Clad Degradaten [

1 BU2 - FCS Leekage SU5- RCS Leekage SUS - Urplanned Loss of M Chsaa or Offste GU3 Unplanned Loss of M Chace or Offate -

                                                                                                                                  -t Communcetons Capetilces                          Communcetons Capabilems SU7 - Unplanned Laes of Requred DC Power         EU2 - Loss of Vital 125 Vot DC Power fr           .

Dumg Cold Shutdown or Refunkng Mode for GEATER Tl%N 15 Mnutes Greater Then 15 Mnutes Shutdomo EAL not currardy addrv=aad by GU1 - Unpienned Loss of Any Functon Needed M.JMARC to Marten Cold Shutdown ,

                  /Jert          SA1 - Loss of M Offete Power and Loss of M       EA1 m=rm Blackout VWe on Shutzbn Chaite AC Power Dunng Cold Shutdown or           Coahng                                            ,

Refunkng Mode ; SA2 - Failurs of Raartnr Protettaan Sptam QA1 - Failure of Atmamatic Reactor Trip instrumentecon to Complete or initiate en i Automecc Reactor Scram Chce e Reactor Protecton System hernet Has Been rvenaana '; and Manuel Scram Was Successfd ; SA3 -Inabibey to Martan Plert n Cold QA2 -inab4y to Morten Plant in Cold i ShtAdown Shutdown

                                                                                                            <                       ?

SA4 - W LDos of Most or M Safety GA3 - Urvismed Loss of Safety System System Annunoston or Mcoton in Control Annunostors With Transert in Prtgess , , Room With Esher [1] e Sgnrficert Transiert in Progress or(2) Compensatory W Mcetars ere thoveilable SAS - AC Power Capsbuy to Essarcel B=== EA2 - Only Ore AC Power Source Aveinable to Reduced to e Sngle Power Source for Greater S% 4kVLemy Buses Then 15 Mnutas Such That Any Addconel Sngle Feiture Would Result in Stamon Beckout Site Emergency SS1 - Loss of M Oneite Pcmer and Loss of M ES1 - Stataan Bet:iout Onsee AC Power to Essercel Busses SS2 - Failure of Reetsor Protecnon Sptam GS1 - Foilurs of BOTH Autnmouc MO Manuel Instrumentstacn to Complete or hitiate en Reactor Trip Autorretac Reaczor Scram Once e Raartne Protectaan Speern Satport Has Been Excmedad and Manuel Scrum Wes TOT Arramaful SS3 - Loss of M Vtal DC Pcmor EA3 - Loss of 125 Voit DC Pcw ad R=tre , Trip ES2 - Loss of M 125 Vot DC n maa i Ceh/ert Cirffs EAL Basis Document G.8 June 15,1993

                                   &NCRAL NOTES FOR EAL TECHNCAL BASIS l

Table G3: Comparison of NUMARC Guiderms to BG&E Cs NUMARC System Matfunction Category Ernergency Dess Generic NUMARC C Cohat Diffs C Site Emergency SS4 - Cortplate loss d Furx: ton Noeded to QS2 - Corrpiets loss of Furr:pon tboded to (Continued) hm&Mae M h h & Marrain fit Shutdm SS5 -loss of Water lmol That Has or Will Q33 - toss of Water twal Mr. Can Unccwor Uncosor Fuoi in the Roactor Vossal Fuoi in the Rosator Vessel SS6 - Instmhty to Mors:nr a Syv6cac Transert ES2 - Loss of All 125 Volt DC Buses in PrD9TSs General Emergency SG1 - Pmionged Loss of AX OffGte Power and EG1 - Pmiorgod Staraon Beckout Pruioryod Loss of AN Or>Gte AC Power SG2 - Fess of tre Rosciar Pmtacoon W OG1 - Fa#urs of DOTH Autarnatic AfD Manuar at Manua! Scrurn was fCT % mc='ul and Ros:: tor Tnp -AtB Extrune Chailonge to the There s Irdceuon of an Extrerne Chelierge to Ahicy to Cool the Core tte Atmicy to Cool the Core 4 l l l l l l Cabert Diffs EAL Basis Document G.9 June 15,1993

GEMERAL NOTES FOR EAL TECHNCAL BASS Table G4: Cu w-ison of NUMARC Guidelines to BG&E Cs NUMARC Rasion Product Bwrier Degradetion Category k Emergency Casa Genanc NUMARC C CeNert Diffs C Unusual Event RJ1 - ANY Loss or ## Potortal Loss of BU1 - Loss OR Potartial Loos of CNrMT Bemer Dortsmore Alert FA1 - ## Loss or ## Potarcel Loss d BA1 - Loss OR Potarcal Loss of EmER Fuel EMER Fuel Qad OR ACS Dad Bemer OR ACS Bamar Site Emergency FS1 - Loss of BOTH Fuel Qad NO RCS BS1 - Loss of Potartialloss d ## Two OR Bamers Potarcel Loss of BOTH Fuel Qad NO RCS OR Potarcel Loss of BTKR Fuel Qad OR RCS and Loss of N# Addoonal Bemer Generel Emergency FG1 -(mes of ## Two Bamers BG1 - Loss of Two Bemers NO Putantal tasa NO of Third Bamor Patercelloss of Thrd Bemer Cabert Orffs EAL Basis Document G:10 June 15,1993

GEfERAL NOTES FDR EAL TECHNICAL BASIS Table G5: Comparison of NUMARC GLidelines to BG&E EALs NUMARC Rssion Product Barvier Degredadon Category EALs Generic NUMARC EAL Cahert Offfs EAL Fuel Qad BarTier Fuel Qad 1 - Cnucal Safety Funcuan Status FCB1 - EWaty Functon Status /Functonni Reco sy Fuel Qad 2 Pnmary Cooiart Actmty Lavut FC83 - Radebon Fuel Dad 3 - Com Ext Thermocouple Roadrgs FCB1 - Sefuty Functon Status /Functoral Anco sy FCO2 - Ternperature Fuel Dad 4 - Roactor Vessel Water taval FCB4 - Reactor Vensul Wetzr Level Fuel Osd 5 - Ccrenmort Radauon Monitonrg FCB3 - Rodauan Fuel Osd 6 - Other (SmSpectric) in6cabons FCBS - TC Judgemore Fuel Qad 7 - Emergency Drector Judgemst FC85 - TC Judgomort RCS Bamer ACS 1 - Cnucal Safety Furx:uon Satus AC81 Safety Furcon Ekatus/Funcuanal Aecoc y ACS 2 - ACS Leak Ama RCO2 - Torrporsture AC84 - Coolart Laaksge ACS 3 - SG Tute Rupture RC82 - Turnporature AC84 - Coolart Lankage RCS 4 - Contarmore Radeuon Morronng ACO3 - Radauon RCS 5 - Other (ste+pocrhc) lndcanors ACB1 - Safety Furmon Status /Furconal Roco*Y AC84 - Cooiart Laakage RCS 6 - Emergency Drector Judgornert ACBS - SEC Judgemart Contsnment Bamar Contamiort 1 - Cntical Safety Fvoon Status 001 - Safety Funcuan Status /Funcmonal Aectwry j i Cortarvnont 2 - Contamurt Pressure 0E5 - Pressure Cormnmort 3 - Cormnmort isowaan Valve 064 - Coolant Lankage Status After Containmort teoisorn Cortarrrert 4 - SG Secondary Sde Rokese 004 - Coolart Lasiusgo Wth Pnmery to Secondary Leaksge Cortarrnert 5 - Sgev6 cart Fie30acove kwyrary 003 - Radston in Corterrnort Cormnmort 6 - Core Ext Trortrocouple 002 -Tempursturn Readergs Corsarmart 7 - Ottwr [SeaSpoofcl ir:1 cations OB1 Safoty Furouon Status /Furconal Rocwory 0B3- Ae3ston Cortamort B - Emergency D sc2cr Judgomort 006 - TC Judgerort Cabert Cirtfs EAL Basis Document G:11 June 15,1993 I

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O l l 1 i I O RADIDACTMTY RELEASE O

RADIDACTMTY RELEASE Eme_rgency Classification Levet UNUSUAL EVENT Applicable Operational Modes: All Calvert Cliffs initiatino Condition: RU1 Unplanned Radioactive Release Exceeding 2 X Tech Spec Umits for AT LEAST 60 Minutes NUMARC Recognition Catacos Abnomial Red Levels / Radiological Effluent NUMARC Initiatino Condition: AU1 Any Unplanned Release of Gaseous or Liquid Radioacivity to the Envimnment that Exceeds Two Times i the Radiological Technical Specifications for 60 Minutes or Longer Barrier Not Applicable NUMARC Generic Basis: Unplanned, as used in this context, includes any release for which a radioactive discharge permit was not prepared, or a release that exceeds the conditions (e p., minimum dilution flow, maximum discharge flow, alarm setpoints, etc.) on the applicable permit. Valid means that a radiation monitor reading has been confirmed by the operators to be cormat. Unplanned mieases in excess of two times the site technical specifications that continue for 60 minutes or longer represent an uncontrolled situation and hence, a potential degradation in the level of safety. The final integrated ( dose (which is very low in the Unusual Event emergency class) is not the primary concem here; it is the degradation in plant control implied by the fact that the miease was not isolated within 60 minutes. Therefore, it is not intended that the release be averaged over 60 minutes. For example, a release of 4 times TS for 30 minutes does not exceed this initiating condstion. Further, the Emergency Diredor should not wait until 60 : minutes has elapsed, but should declare the event as soon as it is detc mined that the release duration has ) exceeded or will likely exceed 60 minutes. l l For sites that have eliminated effluent technical specifications as provided in NRC Generic Letter 89-01, the comesponding maximum limit fmm the site's Offsite Ocse Calculation Manual should be used as the numeric basis ; of the EAL 10 CFR 50.72 mquires a non emergency four hour report for release that exceeds 2 times maximum permissible l concentration [MPC) in unrestricted areas averaged over a period of one hour. There is generally more than one applicable technical specification (e.g. air dose rate, organ dose rate, organ doses, release rate, etc.). Often, effluent monitor alarms are based on instantaneous release rates. Depending on the soume term, other techr.ical specifications may be more limiti ng. For this reason. the EALs should trigger an assessment of all applicable specifications. Monitor indications should be calculated on the basis of the methcdology of the site Offsite Oose Calculation Manual (OOCM), or other site procedures that are used to demonstrate compliance with 10 CFR 20 and/or 10 CFR 50 Appendix I requirements. Annual average meteorology should be used where allowed. <> In < Generic > EAL 3, the 0.10 <rnnem/h> value is based on a pmration of two times the 500 mrem /yr basis of 1 the 10 CFR 20 normccupational <OAC> limits, ruunded down to 0.10 < mrem /h>. If other Site-Specife values are applicable, these should be used. Calvert Cirffs EAL Basis Occument R;1 June 15,1993

RACSACTMTY RELEASE Some sites may find it advantageous to addmss gaseous and liquid releases with separate inidating conditions and EAl s. Plant-Soecife Information: Wth the change in 10 CFR Pan. 20. the term MHC has been superseded by the term DAC [ Derived Air Concentration). The new rule has also reduceo ne narroccupational radiation exposum from 500 mmm/yr to 100 mrem /yr. Calvert Chffs will use the 500 mmm/yr value consistent wth its Technical Specifications. Cabrt Cirffs does not have either a perimeter radiation montoring system or automated ruskime dose assessment capabilty. Thus, the generic EALs mcommended by NUMARC do not apply to the Calvert Ctrffs Nuclear Power Plant. The main plant vents consist of the exhaust flow from the auxiliary building ventilation systems and the condensate offgae system. Batch mleases fmm the Waste Gas Decay Tanks, containment vents and containment purges am also directed into this stream. Per OOCM Attachment 7 the Unit 1 and Unt 2 vent flow rates are assumed to be 59.4 m'/sec and 47.1 m'/sec respectivey Each plant vent is monitored by a beta sensitive plastic I scintillator Wide Range Noble Gas Monitor (WRNGM 1-RIC-5415 and 2-RIC 5415) which is displayed in Ci/sec , and a GeigerWluller tube Mein Vent Montor (1-RL5415 and 2-RL5415) which is displayed in CPM. The values used for the EALs were determined assuming annual average meteomlogy, RCS noble gas concencrations, and using dose conversion factors used for emergency pmparedness off-site dose assessment. The total gaseous miease corresponding to 2 times Technical Specifcation limbs is appmximately 0.114 mrem in one hour, as calculated using the equation below. 2 x Techr.ical S scifcation - 2 x 500 mrum/ year - 1000 mrem / year Hours / year - 24 x 365 - 8760 hours / year (1000 mrem / year] / [8760 hours / year) - O.114 mrem / hour (or 1.14E-3 mSv/ hour) The values for the vent redstion monitor readings are based on 90% of the 2 OAC [ derived air concentration) < at the site boundary. This reduction will account for events that may result in releases through both unit vents. l The 10% factor allowance for the other unit vent is cc.iservative because it is two to three ortiers of magnitude i larger than the normal releases through each vent. For the main vent monitors, which read in CPM, the Une ; 1 flow i ' .s assumed because c +" result in the lowest concentration. ! l

i RC5415 EAL Threshold l Per Reference 5. Tech Spec limit cormsponds to 1.8 E+5 Ci/sec (see tota!) 2 x 1.8 E+5 pCi/sec - 3.6 E+5 Ci/sec Assume event in one unt, allow 1G6 for release fmm other uns l RIC-5415 EAL Thmshold - O.9 x 3.6 E+5 Ci/sec '

                                                       = 324 E+5 Ci/ sec Read as 32 E+5 Ci/sec Catvert Clrffs EAL Basis Document                       R:2                                         June 15,1993

i RADIDACTMTY RELEAT Thus, EAL 1 is wntren as: Valid WRNGM (RICr5415) Reading of AT LEAST 32 E+5 pCi/sec for GREATER THAN 60 Minutes Minimum Conmntraton Conssponding to RL5415 Reading Concentration - Helease rate [uCi/seej Row rate (cc/sec) Unit 1 DOCM flow rates - 59.4 m'/sec Unit 1 Conmntration - 324 E+5 uCi/sec 59.4 m'/see x 10s cc/m'

                                                       - 5.5 E-3 Ci/cc Unrt 2 OOCM flow rates - 47.1 m'/sec Unit 2 Concentration - 3.24 E+5 uCi/_sec 47.1 ma /sec x 10" cc/m'
                                                       - 6.9 E-3 pCi/cc Convert Concentrtrion to CPM for Rb5415 Reading (See Refemnce 5 for Isotooic Mixture) isotope       RCS          % Total        tht i         unt 2            Monitzr       lhti         lht2 Ncontnrjon                  Gws-.         Gws-.               Effoency       CPM         CPM LD/cc)        DD/ct)          (CPM /1D*)

O' Kr@ 5 O.43 9.62 53 E-4 6.6 E-4 35 1.9 E4 23 E4 Kr@ 5 m O.16 3.58 2D E4 2.4 E-4 55 1.1 E4 1.3 E4 Kr4 7 0.15 336 1 B E-4 2.3 E-4 218 4D E4 5D E4 , i Kr48 028 636 3.4 E 4 43 E-4 789 9.8 E4 12 E5 l Xe-133 2.6 58.17 32 E-3 A D E-3 1.87 6.0 E3 7.5 E3 Xe-135 0.85 19D1 1 D E-3 '3r 70 7D E4 9.1 E4

Totals 4.47 100D0 5.5 E-3 6.b 3 3 2.4 E5 3DE5 j l Use krwer CPM va..,a and read as 2D E5 CPM j 1 Thus. EAL 2 is written as: l l Valid Main Vent Monitor (RL5415) Reading of AT LEAST 2D E+5 CPM for GREATER THAN 60 Minutes in a similar manner to that shown for Rb5415. values wem determined for the Waste Processing Monitor (1-RL 5410 and 2-RF5410) assuming noble gas distribution for Wasta Gas Decay Tank rupture, everage annual 8 meteomlogy, and a nominal waste pmcassing ventilation flow of 23.4 m /sec (49,500 CFM). At 2 DAC at the site boundary, per Reference 5 this conssponds to a meding of 4D E+5 CPM. Thus EAL 3 is wn:cen as: Valid Waste Pmcessing Monitor (RL5410) Reading of AT LEAST 4D E+5 CPM for GREATER THAN 60 7 5 Minutes l Calvert Clrffs EAL Basis Document R:3 June 15,1993

RADIOACTMTY FELEAEE in a similar manner to that shown for Rb5415. values were determined for the Fuel Handling Monitor (ORF5420) assuming only monitor response to noble gas mieased fmm a Fuel Handling incident, average annual meteomlogy, and a nominal fuel handiing area ventilation flow of 15.1 m'/sec (32,000 CFM). At 2 DAC at the site boundary, Q per Refemnce 5 this corresponds to a reading of 3.4 E+5 CPM. Thus, EAL 4 is wrcten as: Valid Fuel Handling Monitor (RL5420) Reading of AT LEAST 3.4 E+5 CPM for GREATER THAN 60 Minutes Analysis was also performed for potential mieases through Access Contml Point and ECCS Pump Room. Per Reference 5. for each of these locations 2 DAC at the site boundary comespond to monitor medings that are greater than 1 E+6 CPM, l.a., off-scale high. Thus EAL 5 is written as: Valid Access Contml Monitor (RL5425) Reading Off-Scale HIGH for GREATER THAN 60 Minutes EAL 6 is wrezen as: Valid ECCS PP Room Monitor (RL5406) Reading Off-Scale HIGH for GREATER THAN 60 Minutes Uguid effluent is monitored by the Liquid Weste Dscharge Radiation Monitor (D-R E-2201). A high radiation alarm fmm this monitor will msult in a signal to close the Liquid Waste Dscharge Vatves. If these valves will not shut, the operators will stop the pump being used for the discharge and shut the Liquid Waste RMS Dutlet valve. It is extremely impmbable that a liquid effluent discharge for greater than 60 minutes could exist following a valid monitor alarm. Thus, no EAL for liquid effluent release is required. Source Documents / References / Calculations: , t

1. Technical Specif~c ations

TS 3/4.32 Monitoring instnxnentadon - Radiadon Monitoring instnamentation
TS 3/4.11. Radioactue Effluents

2. Abnormal Operating Pmeedures

ADF6B Accidental Release of Radioactive Liquid Waste

3. System Descriptions

No.15. Radiation Monitoring System

4. Off Sea Dose Calculadon Manual (ODCM) for the Baltimore Gas & Electric Company Ce3 vert Chffs Nuclear Power Plant

5. Radoactivty Release Emergency Action Levels, J.B. Mcitvaine, JS8 Assocates, Inc., September 1990

6. 10 CFR Part 20. Standartis for Protection Against Radiation; Rnal Rule,56 FR 23360, May 21,1991 O

Calvert Cliffs EAL Basis Document R.4 June 15,1993

RADIDACTMTY ELEAEE . 1 Emeroency Classifcation Levet UNUEiUAL EVENI' eppleable Doerational Modes: All 1

 - Calvert Cliffs initiatino Condition:                                                                                     l RU2 Unaq=*=i incmase in Plare Radiabon NUMARC Recognition Cateoorv Abnormal Red Levels /Radiolog' cal Effluent i

NUMARC Initiatino Conditi_ ort: ; AU2 Unexpected increase in Plant Radiation < > Barrier.' Not Applicable .j NUMARC Generic Basis: Valid means that a radiation monitor reading has been confirmed by the operatom to be cormet. ,

  < Events associated with this IC> tend to have long lead times relative to potential for radiological release outside the site boundary; thus, impact to public health and safety is very low. < >                                              ;

i in light of Reactor Cavity Seal failure incidents at two different PWRs occurring since 1984, explicit coverage of l these types of events via EALs 1 < indication of ur, controlled water level decrease in the reactor refueling cavity ; with all irradiated fuel assemblies remaining water covered > and 2 <indicaton of uncontrolled water level j decrease in the spent fuel and fuel transfer canal with all irredeted fuel assembles remaining water covemd> ; is appropriate given their potential for incmased doses to plant staff. Classifcation as an Unusual Event is l warranted as a precursor to a more serious event. < > { EAL <2 (valid Direct Area Radiation Monitor readings increase by a factor of 1000 over normal levels whers Normal levels can be considemd as the hyhest rsading in the past twentyfour hours excluding the current peak i valueP addresses unplanned increases of ir> plant radiation levels that represent a degradaten in the control of radioactue material, and represent a potential degradation in the level of safety of the plant. This <lC> escalates - t to an Alert per <C RA2, Damage DR Uncovery of irradiated Fuel Dutside the Reactor Vessel, or C RA3, Radiation i increases That Impede Safe Operation >. ' Plant 4pecific information: EALs m!ated to dry storage of spent fuel in Horizontal Storage Modules are separately addressed under RU3, i Potential Degradation of Containment of Dry Stored Spent Fuel j Of concem in this C are water level decreases over spent fuel that are suffcient to cause noticeable incmases in measured radiation levels. Additionally, fuel handling incidents can lead to many of the same symptoms of ; increased plant radiation levels. Existence of spent fuel pool alarm (RF5420), spent fuel pool area monitor alarm ' j (Rh7024), or containment monitor (RL5316A/B/C/D) reading of at least 100 mrsm/h is used as the thmshold for entry into this C. One hundred mrern/h corresponds to the administretwe limit for a high radiation area and ' is signifcantly higher than the dose rates expected for fuel handling. Thus, EAL 1 is wntzen as: ADRED Dr ADP4E is implemented AND Any of the Following:

Valid Spent Fuel Pool Alarrn (RL5420) !
Valid Spent Fuel Pool Area Monitor Alarm (Rh7024)
Valid Containment Radiation Monitor (RF5316A/B/C/D) Reading of AT LEAST 100 mrsm/h t

I Catvert Cirffs EAL Basis Document R.5 June _15,1993

RADIDACTMTY FELEASE i EAL 2 is taken dimctly fmm NUMARC. Momentary incmases due to events such as msin transfers or contmlied I movement of redoactrve soumes should not msult in emergency declaration. Thus, a threshold of 5 minutes is selected to pmclude such spikes. Certain radiation monitor alarms may go offscale high before maching 1000 times normal readings. s Thus, EAL 2 is written as: Valid Unexpected Rad Monitor Reading Offscale High OR GREATER THAN 1000 Times Normal Reading for GREATER THAN 5 Minutes j Valid means that the indcation is from instrumentation deterTnined to be opemble in accordance with the Technical Specifcations or has been venfied by other independent methods such as indications displayed on the contml panels, reports fmm plant personnel, or radiological survey results. The Unusual Event may be terminated when the following actions occur: (1) The source has been identifed, '" (2) The source has been contmiled or contained as appropnate, and (3) Appmpriate personnel radiation practices have been implemented. Expected increases in radiation monitor readings due to controlled evolutions (such as lifting the mactor vercal head during refueling) should not result in emergency declaration. Irt-plant radiation level increases that wouti result in emergency declaration am also unplannect e.g., outside the limts established by an existing radioactive discharge permit. Sourte_Dgguments/_Referencesf_ Calculations:

1. System Descriptions

No. 10. Spent Fuel Pool and Spent Fuel Pool Cooling And Purifcation Systems
No.13. Refueling Equipment
No. 15, Radiation Monitoring System

2. Abnormal Operating Pmcedures

ADP-60. Fuel Handling incident
ADPEE, Loss of Refueling Pool Level Calvert Clrtfs EAL Basis Document R.6 June 15,1993

I RADOACTMTY RELEAE i Ememency Classification Level: UNUSUAL EVENT ; Appigb_le_O_perational Modes: AU. OL l Catvert Cidfs initiatino Conde, ion: RU3 Potantaal Degradation of Containment of Dry Stored Spent Fuel h!UMARC Recoonition Catego_ry Abnonnal Rad Levels / Radiological Effluent NUMARC initiatino Conde, ion: AU2 Unexpected incmase in Plant Radiacion < > l Barrier: Not Applicable ]a l NUMARC Generic Basis- 1 Vahd means that a radiation monitor reading has been confirmed by the operators to be cormat. l 4 vents associated with this IC> tend to have long lead times m!stive to potential for radiological mlease outside the site boundary, thus, impact to public health and safety is very low. o

  <This IC> applies to plants with licensed dry storage of older irradiated spent fuel to addmss degradation of this spent fuel. One utility uses values of 2 R/hr at the face of any dry storage modue or 1 R/hr one foot away fmm a damaged module. o l

Plant-Specific information: As a result of a meeting between BG&E Emergency Planning Staff and NRC Facilities Radiological Safety and l Safeguards Branch personnel on July 18,1992, the following EALs wem developed regarding potential degradation of containment of dry stomd spent fuel. 1 EAL 1 is wntten as- ! I Horizontal Storage Module (HSM) Access Door Contact Dose Rate GREATER THAN 500 mrem /h At 100 mm per Sievert, this corresponds to e dose rate of 5 (milliSeverts) mSv/h. l EAL 2 is written es: l Horizontal Storage Module (HSM) Side Walt Door Contact Dose Rate GREATER THAN 100 mrem /h i This cormsponds to a dose rate of 1 rrSv/h. ] EAL 3 is wntzen as: l Any Unplanned Event Outside the Auxiliary Building Resulting in the Smach of a Dry Shielded Canister l (DSC) Containing Spent Fuel Unplanned is used to preclude declaration of an emerDency whem the DSC has been intentionally opened for maintenance or mpair actMty in accortlance with a valid radiation work permit. l

%                                                                                                                     i l

l l l l June 15,1993 l Calvert Cirffs EAL Basis Document R:7

RADOACTMTY FELEASE EAL 4 is wrcten as: Dry Shielded Canister (DSC) Transfer Cask Containing Spent Fuel Has Been Dmpped from the Trailer Soume Documents / References / Calculations: ,

1. Letter dated September 24.1992. LB. Russell (BG&E) to US Nuclear Regulatory Commission, re:

Emergency Action Level Review Meeting held on JuV 16,1992

2. BG EC1-121. Rev. O, An Assessment of Storage Term Radiological Exposure Rates at the Calvert Cirffs NUHOMSc ISFSI. Pacific Nuclear Fuel Services. Inc., September 1990 O

i 1 i l l l i I O CaNert Cirffs EAL Basis Document R.B June 15,1993

RAOlGACTMTY RELEASE Emeroency Classifcadon Leve_l: ALERT p Applicable Operationai ;/,~ier All

L ]\

Calvert Cirffs InidatLng Condfio_.n: RA1 Unplanned Radioactive Release Exceeding 200 X Tech Spec Umits for AT LEAST 15 Minites NUMARC Reccancion Catego_ry Abnorma! Rad Levels /Radiologcal Effluent NUMARC intiating Condtion: AA1 Any Unplanned Release of Gaseous or Uguid Radcactkity to the Envimnment that Exceeds 200 Times Radiological Technical Specifcadons for 15 Minutes or LonDer _Barner- Not Applicable NUMARC Generic Basis: Valid means that a radiation mon't or mading has been confirmed by the operators to be correct. This event escalates fmm the Unusual Event by escalating the magnitude of the release by a factor of 100. Prorating the 500 < mrem /yr> cnterion for both time (8766 hr/yr) and the 200 multiplier, the assocated see boundary dose rate would be 10 <mmm/h>. The required release duration was reduced to 15 minutes in recognition of the increased seventy.

/N   For sces that have eliminated effluent technical specifcadons as prtmded in NRC Genenc Letter 8901, the corresponding maximum hmt from the sce's Offste Oose Calculation Manual, muldplied by 200, should be used as the numeric basis of this EAL Monitor indications should be calculated on the basis of the methodology of the site Offsite Oose Calculadon Manual (DOCM), or other site procedures that are used to demonstrate compliance with 10 CFR 20 and/or 10 CFR 50 Appendix I requirements - adjusted upwanis by a factor of 200. Annual average meteorology should be used where allowed. < >

In <Genenc> EAL 3, the 10 <mmm/h> value is based on a proradon of 200 times the 500 mrem /yr basis of the 10 CFR 20 non-occupadonal MPC limits, rounded down to 10 < mrem /h>. If other Ste-Specific values are applicable, these should be used. P_lant-Specific Information: Wth the change in 10 CFR Part 20, the term MPC has been superseded by the term OAC (Derked Air Concentration). The new rule has also reduced the nonoccupational radiadon exposure from 500 mrem /yr to 100 mrem /yr. Catve.t Cliffs will use the 500 mrem /yr value consistent wth its Technical Specifcadons. Cakert Cliffs does not have ether a penmeter radiadon monitoring system or automated reaktime dose assessment capability. Thus, the genenc EALs recommended by NUMARC do not apply to the Calvert Cliffs Nuclear Power Plant. A description of the applicable monitors and the methods used to calculate EAL values is shown in RU1, Unplanned Radioac;ve Release Exceeding 2 X Tech Spec Limits for GREATER THAN 60 Minutes. Values foe this IC are based on tb values shown in RU1 muldplied by 100. (See equation below] Nus ] - Cabert Cittfs EAL Basis Document p9 June 15,1993

RADIGACTMTY RELEASE RA1 Thmshold for RC5415. RL5415 RU1 Value x 100 - RA1 Value For RG5415 32 E+5 Ci/second x 100 - 32 E+7 Ci/second For RL5415 2.0 E+5 CPM x 100 - 2.0 E+7 CPM (Above top of scale) The ECCS PP Room Monitors (1/2-RL5406) and the Access Control Monitor (0 Ab5425) am not considemd ' hem because they will be offscale high at the Unusual Event emergency classification level. At the Alert level, the readings on the main vent monitors (1/2-Af 5415), the Waste Pmeessing Vent Monitors (1/2-RL5410), and the Fuel Handling Area Vent Monitor (OLakS420) cormspond to medings well above the top of the range (1.0 E+B CPM) for these instruments. Therefore, these monitors pmvde no useful information for this IC and are excluded fmm consideration. Thus, EAL 1 is wntten as: 1 Valid WRNGM (RIC-5415) Reading of AT LEAST 32 E+7 Ci/see for GREATER THAN 15 Minutes The purpose of the Main Steam Effluent Radiation Monitor Systern is to monitor possible noble gas releases to the atmosphem from the main steam line through the atmospheric steam dump vakes, the main steam safety relief vakes, and the auxiliary feedwater staam turbine exhaust. The system includes two redstion monitors (1/2-Rh5421 and 1/2-Rh5421) for each unit - one radiation monitor for each steam generator. Each radiation ' detector is an ion chamber filled with xenon gas with a small " keep alive

sourte that produces a reading on the corresponding main control board rate meter of about 10* R/hr.

The noble gas miease rate of 32 E+7 Ci/second (which corTesponds to a whole body dose of 10 mrem in one hour at the site boundary) may also occur thmugh release via main steam safety valve or atmospheric dump vake. By reverse calculatiori using Attachment 3 of ERPIP 821 (see box below): O Cakert Cirffs EAL Basis Document R:10 June 15,1993

RADIQACTMTY FELEAEE l RA1 Threshold for RL5421, RL5422 Release Rate - 32 E+7 pCi/sec (see above) Release Coeffcient (for SG Tube Ruptum) - 6.1 E+2 pCi/cm' rem /h Atmospheric Dump Valve Flow Rate - 1.4 E+6 cm'/sec Safety Valve Flow Rate - 2.4 E+6 cm'/see Main Steam Monitor Reading (rem /h) - Release Rate Release Coefficient x Flow Rate For safety valve rum /h = 32 E+7 6.1 E+2 x 2.4 E+6

                                                               - D22 rem /h (read as D2)(O2 mSv/h)

For atmospheric dump vatve mm/h - 32 E+7 6.1 E+2 x 1.4 E+6

                                                               - D38 rem /h (read as D4)[0.4 mSv/h)

The minimum reading for RL5421/5422 is 10 mrem /h due to the " keep alive" soume. Twenty mrem /h would be diffeult to med accurataty. The high alarm setpoint for these monitors is set at 47 mrem /h 5 mrem /h. Therefore, for human factors masons, the existence of the high alarm setpoint is used as the thmshold for this \ EAL Thus, EAL 2 uses the lower value and is written as: Valid Main Steam Effluent Monitor (RL5421, RF5422) High Alarm for GREATER THAN 15 Minutes Valid means that the indication is fmm instrumentation determined to be operable in accordance with the Technical Specifcations or has been verthed by other independent methods such as indcations displayed on the contml panels mports fmm plant personnel, or radological survey results. Based on the March 14,1993 SG tube rupture event at Palo Verde Unit 2 the main steam effluent monitors (Rh5421. RL5422) may read N'8 immediately following SG tube ruptum and prior to reactor trip. However, given the short half 4rfe of N. this should clear within the first minute following reactor trip. Source Documents / References / Calculations:

1. System Descriptions

No. 15. Radiation Monitoring System

2. Off-Srte Dose Calculation Manual (DOCM) for the Baltimom Gas & Electric Company Calvert Cliffs Nuclear Power Rant

3. RadoactMey Release Emergency Anion Levels, J.B. McIlvaine, JSB Associates. Inc., September 1990

4. Emergency Response Ran implementation Procedums

ERPiP 821. Accidental Radoactivity Release Monitoring and Sampling Methods

\ Calvert Cliffs EAL Basis Document R:11 June 15,1993

RADGACTMTY RELEASE l

5. BG&E Intemal Memorandum, J. R. Hill (Nuclear Rant Operations) to R. L Wendedich. CE Operations Subcommittae Meeting - Trip Report, April 16,1993

6. 10 CFR Part 20, Standards for Pmtecten Against Radiation; Rnal Rule,56 FR 23360. May 21.1991

7. Calvert Chffs instructaons

CCl-302, Cakert Cirffs Alarin Manual. Main Staam Effi Rad Monitor 2C24B O

O Cabert Clifts EAL Basis Document R:12 June 15,1993

RADOACIMTY RELEASE lEmemency Classication Level ALERT Applicable Doerational Modes: All QaNert Chffs initiating Conddo_rl: RA2 Demage OR Uncovery of Single krediated RJet Assembly Outside the Reactor Vessel NUMARC Recoonition Categog Abnormal Rad Levels /Pedological Effluent NUMARC Initiatino Conddon: AA2 Major Damage to irradiated Fuel or Loss of Water Level that Has or Will Result in the Uncovering of irradiated Fuel Dutside the Reactor Vessel Darvier Not Apploable NUMARC_ Generic Basis: This IC applies to spent fuel requiring water coverage anr' is not intended to addmss spent fuel which is licensed for dry storage, which is discussed in <RU3, Potentaal Degradation of Containment of Dry Stored Spent Fue>. NUREGO818. Emergency Action levels for Light Water Reactors, forms the basis for these EAls. Each site should also define its EALs t,y the specife area where Irradiated fuel is located such as Reactor Cavity, Reactor Vessel, or Spent Fuel Pool There is time available to take correctke actions, and there is lule patential for substantial fuel damage. In ('_ addLion, NUREG/CRA982. Sevem Accident in Spent fvel Pools in Suppon; of Generic Safuty issue B2, July 1987, indicates that even if corrective actions are not taken, no prompt fatalcies are predicted, and that risk of injury is low. In addition, NRC information Notice No. 9008, KRE5 Hazards Imm Deceped Fuel, presents the following in es discussion: In the event of a serious accident invoMng decayed spent fuel, protectke actions would be needed for personnel on site, while off site doses (assuming an exclusion ama radius of one mile fmm the plant site) would be well below the Environmental Protection Agency's Pmtacthe Action Guides. Accortlingy, it is important to be able to propert/ survey and montor for Kr45 in the event of an accident with decayed spent fuel Licensees may wish to reevaluate whether Emergency Action Levels specifsd in the emergency plan and pmcedures goveming decayed fuethandling actmties appropriateh focus on concem for onsite workers and Kr45 releases in areas where decayed spent fuel accidents could occur, for example, the spent fuel pool working floor. Furthermore, Icensees may wish to determine if emergency plans and corresponding implementing pmcedures address the means for limiting radiological exposures of onsite personnel who are ; in other areas of the plant. Among other things, moving oraite personnel away from the plume and shutting l off building air intakes downwind fmm the source may be appropriate. j i Thus, an Alert Classifcation for this event is appropriate. Escalation, if appropriate, would occur via <other i Radioactuty Release ICs or SEC Judgement Cs>.

BankSpecific information: ADP-6E Loss of Refueling Pool Leve!, provides actions to respond to a loss of Refueling Pool (RFP) inventory due to failure of the Refueling Pool Seal, Steam Generator Nozzle Dams, or the Refueling Pool Drain Line. These actions include p! acing spent fuel in safe storage locations (i.e., all spent fuel will remain water covered following ( pool draindown to the reactor vessel f!ange elevation). If any spent fuel assemby can NOT be placed in an appropriate safe storage location, this corresponds to entry into this C. Cakert Chffs EAL Basis Document R:13 June 15,1993

RADIGACTMTY RELEASE Thus, EAL 1 is wntran as: ADPEE, Loss of Refueling Pool Level, is implemented AND Valid Containment Radiation Alarm (RF 5316A/B/C/D) EAL 2 is wntten as: ADPED, Fuel Handling incident, is implemented AND Any of the Following:

Valid Containment Radiation Alarm (RF5316A/B/C/D) ~
Valid Spent Fuel Pool Radiation Monitor (RL5420) Reading of AT LEAST 2E4 CPM
Valid Spent Fuel Service Platform Monitor (RL7025) Reading of AT LEAST 100 mrem /h Valid means that the indication is from instrumentation determined to be operable in accordance with the Technical Specifications or has been venfied by other independent methods such as indications displayed on the contml panels, Figure R1: RI-5420 Reading for reports fmm plant. personnel, or Single Fuei Rod Damage vs Time radiological survey msults SWe l The containment radiation alarm corresponds to a dose rate of 120E+05 -

200 mmm/h. ,1.00E+05 -

                                          ,e 8.00c+04 -

The value for RF5420 was j 6.00E+04 - determined based on a fuel y 4.00E+04- l handling accident damaging one 2.00E+04 fuel rod in an average (unpeaked) 0.00E+00 7 720 7200 72000 { g fuel assembly. The results of the calculation, showing RL5420 72 Time (Hrs) After Shutdown msponse versus age of the assembly (time after shutdown),is shown as Figure R1. The value of 2E4 CPM corresponds to the minimum expected response and is significantJy higher than the alarm setpoint of 600 CPM. One hundred mrum/h is used for the Service Platform Monitor (RF7025) because it. cormsponds to the i administratke limt for a high radiation area and is signifcantly higher than the dose rates expected for fuel handling actrvites. l Expected increases in monitor readings due to controlled evolutions (such as lifting the reactor vessel head during refueling) should not msult in emergency declaration. Nor should momentary increases due to events such as resin transfers or contro!!ed movement of radioacche sources result in emergency declaration. in-plant radiation level increases that would result in emergency declaration are also unplanneci e p., outside the limits established by an existing radioactive discharge permit. i Source Documents / References / Calculations:

1. System Descriptions

No.15. Radiation Monitoring System

2. Abnormal Operating Procedures

ADPED, Fuel Handling incident
ADP6E, Loss of Refueling Pool Level

( 3. Ogden Calculation #RA-1. ORL5420 Detector Response to Fuel Handling Accxient Calvert Chffs EAL Basis Document R:14 June 15,1993

RADOACTMTY FElEASE Ememency Classifcation Level: ALERT ! Applicable Operational Modes: All Cabert Cliffs initiatino Condcion: RA3 Radiadon increases That impede Safe Rant Opersbon MJMARC Rec _ogntion Cat _ ego _ry Abnormal Rad Levels / Radiological Effluent NUMARC inciatino Condition: AA3 Release of Radioactive Material or incmases in Radiatson Levels Wthin the Facilty That impedes Operation of Systems Requimd to Maintain Safe Operations or to Establish or to Maintain Cold Shutdown Barrier Not Applicable NUMARC Generic Basls: Valid means that a radiation monitor reading has been confmed by the operators to be corvect. This 0 addresses increased radiation levels that impede necessary access to operating stations, or other areas containing equipment that must be operated manually, in order to maintain safe operation or perforrn a safe shutdown. It is this impaired ability to operate the plant that results in the actual or potential substantial degradation of the level of safety of the plant. The cause and/or magnitude of the increase in radiation levels is not the concem of this C. The <5EC> must consider the sourt:e or cause of the increased radiation levels and determine if any other C may be involved. For example, a dose race of 15 < mrem /h> in the mntre' room may ( . be a problem in itself. However, the increase may also indicate high dose rates in the containment due to a LOCA. In this latter case, SAE or GE may be indicated by fission product bemer matrix Cs. At mutiple unit stes, the example EALs could result in declaration of an Alert at one unit due to a radioactivity increase or radiation shine resulting fmm a major accident at the other unit. This is appropriate if the increase impairs operations at the operating unit. This 0 is not meant to apply to increases in containment dome radiation montors as these events <> are addressed in the fission product bamer matrix Cs. Nor is it intended to apply to anticipated temporary increases due to planned events (e.g., irx: ore detector movement, radwaste container movement, depleted resin transfers, etc.) Emergency planners developing the (Ste6pecific) lists may refer to the site's abnormal operating pmcedures, emergency operating procedures, the 10 CFR 50 Appendix R analysis, and/or, the < analysis > performed in response to Section 2.1.6b of NUREGO578, TM12 Lassons Laamed Task Force Status Report andShort-Term Recommendations,when identifying amas containing safe shutdown equipment. Wth regard to the NUREGO578 analysis, do not use the dose rata postulated therein as a basis for the radiation monitor readings for this IC, as the NUREGO578 < analysis > address general emergency mnditions. Areas requiring continuous occupancy include the control room and, as appmpriate to the site, any other control stations that are manned continuously, such as the radwasta control room or a central security alarm station. The value of 15 mmm/h is denved from the GOC 19 value of 5 rem in 30 days with adjustment for expected occupancy times. Although Section 111.0.3 of NUREGO737,Clarifcatson of TMi Action Plan Requirements, provides that the 15 < mrem /b> value can be averaged over 30 days, the value used here is without everaging, as a 30 day duration implies an event potentially more signifcant than an Alert. For areas requiring infrequent access, the (Site-Specific) value(s) should be based on radiation levels which result in exposure control measures intended to maintain doses within normal occupational exposure guidelines and limits (i.e.,10 CFR 20), and in doing so, will impede necessary acx:ess. For many areas, it. may be possible to Cabert Cirffs EAL Basis Document R:15 June 15,1993

RADIDACTMTY RELEAEE establish a single < Generic > EAL that represents a mutiple of the normal radiation levels (e.g.,1000 times normaQ. However, amas that have nommily high dose rates may require a lower mutiple (e.g 10 times normaQ. Sant40ecific Information The control room is required to be continuously occupied following design basis accidents. All actions required to achieve and maintain cold shutdown can be accomplished fmm the control room. Postaccident doses have been evaluated and shown to be less than limits based on GDC 19. On a control room high radiation signal, the l contml mom emergency ventilation system automatcalty swtches into a rectroulation mode of operation with flow through the HEPA fiters and charcoal absorber banks. EAL 1 is based on the GDC 19 limit recommended by NUMARC- ! Thus, EAL 1 is wrcten as: l Valid Contml Room Radiation Monitor (RI-5350) Reading GREATER THAN 15 mrem /h l This corvesponds to a dose rate of 0.15 mSv/h. Valid means that the indication is from instrumentation determined to be operable in accordance with the Technical Specrfications or has been venfied by other l independent methods such as indications displayed on the contml panels, reports from plant personnel, or i radiological survey results, j EAL 2 addresses event sequences outside the plant design basis. In accordance with ERPIP 832. Emergency Work Permits, entry into any area with exposum retes of at least 250 R/h (2.5 Gmy/h) is pmhibited for plant ; saving missions. Thus, EAL 2 is written as: Exposum Rate of AT LEAST 250 R/h in Areas Required to Achieve or Maintain Safe Shutdown l Required means that entry into the ama is not optional and is imperstke based on existing conditions. Areas of concem for Safe Shutdown are listed below. Areas of Concem for Safe Shutdown

Contml Room

Electrical Penetration Rooms

Contml Room HVAC Room

Auxiliary Feedwater Pump Room

Cable Spreading Room

Charging Pump Rooms

Cable Chases

Desel Generator Rooms

Switchgear Room

Refueling Water Tank (RWT) 11(21)

ECCS Pump Room

Condensate Storage Tank (CST) 12

Service Water Pump Room

Pretreated Water Storage Tank (PWST) 11(21)

Component Cooling Pump Room

Fuel Oil Storage Tank (FDST) 12

Main Steam Penetrwtion Room This list of Safe Shutdown areas is displayed on the EAL Tables to assure that all areas related to Safe Shutdown are considered by the SEC.
Expected increases in monitor readings due to controlled evolutions (such as lifting the reactor vessel head during refueling) do not resut in emergency declaration. Nor should momentary increases due to events such as resin transfers or controlled movement of radoactke sources resut in emergency declaration. Irmlant radiation level increases that would resut in emergency declaration are also unplanned, a g, outside the limits established by an existing radioactive discharge permit. The containment radiation monitor readings should only apply to this IC when personnel are in containment for normal maintenance, inspection, surveillance, testing, or refueling activities. k Calvert Cirffs EAL Basis Document R:16 June 15,1993
RADIDACTMTY PELEASE Soume Documents /Refemnces/ Calculations:
~~1. System Desenptions~~
~~No. 15. Radiation Monitoring System~~
2. Letter, G.C. Cmel (BG&E) to NRC Document Control Desk dated September 1,1989. Contml Room Dose

3. Letter, JA Tieman (BG&E) to AC. Thadani (NRC) dated Mamh 5,1986. Control Room Dose

4. Emergency Response Plan implementation Procedum
~~ERPIP-832. Emergency Work Permits~~
~~5. CCH3OD, Calvert Cliffs Radiation Safety Manual O~~
O Calvert Cirffs EAL Basis Document R:17 June 15,1993 1
RADIDACTMTY RELEASE Ememency Classification Level: SITE EMERGENCY Anoicable Doerational Modes: All _Calv_eri Ciffs initiatina Conditiori: RS1 Off-Sita Dose of AT LEAST O,1 Rem [EDE + CEDE) Or O.5 Rem CDE 1hyroid NUMARC Recognition Catego_ry Abnormal Rad Levels /Radclogical Effluent NUMARQ lnitiatino Condition: AS1 See Boundary Dose Resutinp fmm an Actual or imminent Release of Gaseous Radioactuty Exceeds 100
~~< mrem > Whole Body or 500 < mrem > Thyroid for the Actual or Pmiected Duration of the Release Barrier Not Appicable em NUMARC Generic Basis:~~
Valid means that a rediation monitor reading has been confirTned by the operators to be cormct. The 100 <mrum> integrated dose in this initiating condtion is based on the pmposed 10 CFR 20 annual average population exposure. This value also pmvides a desirable gradient (one order of magnitude) between the Alert, See <E>mergency, and General Emergency classes. It is deemed that exposures less than this limit are not consistent with the Site <E>mergency class description. The 500 <mmm> integrated < > thyroid dose was established in consideration of the 1:5 retic of the EPA Pmtective Action Guidelines for whole body and thymid. Integrated doses are generally not monitomd in real-time. In establishing the emergency action levels, it is suggested that a duration of one hour be assumed, and that the EALs be based on a site boundary dose of 100
< mrem /h> whole body or <500 mrem /h> child thyroid, whichever is more limiting (depending on source term assumptions). If individual site analyses indicate a longer or shorter duration for the period in which the substantial portion of the activey is released, these dose rates should be a4usted.
The FSAR soume terms applicable to each monitomd pathway should be used in conjunction with annual average meteorology in detemiining indcations for the monitors on that pathway. Plant-Specife Infortnation 10 CFR Part 20 was revised following the development of the NUMARC methodology. Catvert Cliffs uses the new rule as its basis for determining dose. Emis the Effecdve Dose Equivalent as defined in 10CFR20.1003. mm is the Committed Effective Oose Equivalent as defined in 10CFR20.1003. CDEis the Committed Dose Equivalent as defined in 10CFP20.1003. Calvert Cliffs does not have either a perimeter radiation monitoring system or automated realtime dose assessment capabiley. Thus, the generic EALs recommended by NUMARC do not apply to the Calvert Cirffs Nuclear Power Plant. A description of the applicable monitors and the methods used to calcukne EAL values for the WRNGM is shown in RU1, Unplanned Radioactive Release Exceeding 2 X Tech Spec Umits for GREATER THAN 60 Minutes. Values for this IC are based on the values shown in RU1 scaled up fmm O.114 mrem in an hour (i.e., hourty rate resulting in 2 X 500 mrem in one year) to 100 mrem (EDE + CEDE) (1 rnSV) in en hour. (See box below.) EALs 1 and 2 only apply if dose assessment capability is not available, i.e., without dose assessment. The preferred method of declaration is via EAL3, wth EALs 1 and 2 as backup methods, if requimd. Calvert Ctrffs EAL Basis Occument R:18 June 15,1993
RADIDACTMIY RELEASE RS1 Threshold for RC5415 m Scale up fmm RU1 uncormcted release rate of 3.6 E+5 Ci/sec RS1 Value - 100 mmm/h x 3.6 E+5 pCi/sec l O.114 mrem /h (or .00114 mSv/h)
~~- 32 E+8 pCi/sec Read as 3 E+8 Ci/sec Thus, EAL 1 is wntten as:~~
Valid WRNGM (RIC-5415) Reading of AT LEAST 3 E+8 Ci/sec for GREATER THAN 15 Minutes (Wehout Dose Assessment) i l This value corresponds to a concentration of about 5 Ci/cc and falls well within the range of the WRNGM. The purpose of the Main Steam Effluent Radiation Monitor System is to monitor possible noble gas releases to , the atmosphem fmm the main steam line thmugh the atmospheric steam dump vanes, the main steam safety miief valves, and the auxiliary feedwater steam tuttine exhaust. The system includes two radiation monitors (1/2-Rh5421 and 1/2-RL5421) for each unit - one radiation rnonitor for each steam generator. Each redation detector is an ion chamber filled with xenon gas wth a small *ksep alive" soume that pmduces a reading on the corresponding main contml board rate meter of about 10' R/hr. The noble gas release rate of 32 E+8 Ci/sec (which corresponds a whole body dose of 100 mrem in one hour at the site boundary) may also occur thmugh release via main steam safety vake or atmospheric dump vake. [] V i i f ( Cakert Cliffs EAL Basis Document R:19 June 15,1993 i
RADIGACTMTY FELEASE I l By reverse wcu6aton using Atr.achment 3 of ERPlP 821: 1 RS1 Thmshold for Ab5421. RL5422 Release Rate - 32 E+8 Ci/sec (see above) Release Coeffcent (for SG Tube Ruptum) - 6.1 E+2 Ci/cm rem /h ! Atmospheric Dump Valve now Rate - 1.4 E+6 cm'/sec Safety Vabe Flow Rate - 2.4 E+6 cm'/sec Main Steam Monitor Reading (rem /h) - Re' esse Rate Release Coefficient x Flow Rate For safety valve mm/h = 32 E+8 6.1 E+ 2 x 2.4 E+ 6
~~- 022 rem /h (read as 0.2)(2 mSv/h)~~
~~For atmospheric dump valve mm/h - 32 E+8 6.1 E+2 x 1.4 E+6~~
~~- O.38 rem /h (read as O.4) (4 mSv/h)~~
Thus, EAL 2 is wntren as: Valid Main Steam Effluent Monitor (Rb5421, RL5422) Reading of AT LEAST 02 rem /h for GREATER THAN 15 Minutes (Wehout Dose Assessment) Valid means that the indication is from instrumentation determined to be operable in accordance with the Technical Specifications or has been venfed by other independent methods such as indcations displayed on the control panels, reports fmm plant personnel, or radiological survey results. Based on tre March 14,1993 SG tube ruptum event at Palo Verde Unit 2. the main steam effluent monitors (Rh5421. RF5422) may read N immediately following SG tube rupture and prior to reactor trip. However, given the short half-Irfe of N'*, this should clear wthin the first minute following reactor trip. Dose consequences can be determined by use of ERPIP 822, Ir>ial Dose Assessment Manual Calculation Methods, or by use of ERPIP 823, Dose Assessment Computer. Dose assessment will be performed in accordance with the new 10 CFR 20 scheduled to take effect January 1,1994. Thus. EAL 3 is wroten as: Dose Assessment Determines integrated Accxient Forecast Dose Off6ite is AT LEAST O.1 rem (EDE + CEDE) Or O.5 rem CDE Thymid This cormuponds to doses of 1 mSv (EDE+ CEDE) and 5 mSv CDE Thymid, respectively. Calvert Clffs EAL Basis Document spo June 15,1993
RAD 13ACTMTY FElEASE Source Documents / References / Calculations:
~~1. System Descriptions~~
^
~~Radiation Monitoring System I~~
~~2. Emergency Response Ran implementation Pmcedums~~
ERPIP-810, Main Steam Radioactroty Release Estimate

ERPIN21, Accidental Radioactkity Release Monitonng and Sampling Methods

ERPIP-822, initial Dose Assessment Manual Calculation Methods j

ERPIP-823, Dose Assessment Computer
3. Radioactkty Release Emeq;ency Action Laels, J.B. Mcitveine, JS8 Associates, Inc., September 1990

4. BG&E Fuel Degradation EALs Calculation Wortsheet, JSB Aswiens, February 18,1993

5. Radioactivity Release Emergency Action Levels, J.B. Mcikaine, JSB Associates, Inc., September 1990

6. BGSE Intemal Memorandum, J. R. Hill (Nuclear Plant Operations) to R. L Wenderlich. CE Operations Subcommittee Meeting - Trip Report, April 16,1993

7. 10 CFR 20, Standards for Protection Against Radiation; Final Rule, SG FR 23360, May 21,1991 0
\ Cakert Cirffs EAL Basis Document R21 June 15,1993 f
RAD E M FELEASE __ i Emeroenev Cassification Level: GEMRAL EMERGENCY i Apolcable Ooerstional Modes: AR l Calvert Cliffs initiatino Condition: ; i RG1 Off-Site Dose of AT LEAST 1 Rem (EDE + CEDE) Or 5 Rem CDE Thymid ; NUMARC Recoonition Cateoorv Abnormal Rad Levels / Radiological Effluent i NUMARC Initiatino Condtion: , AG1 Sce Boundary Dose Resulting from an Actual or imminent Release of Gaseous Radoactivity that Exceeds 1000 < mrem > Whole Body or 5000 <mrerrP Child Thyroid for the Actual or Pmjected Duration of the ! Release Using Actual Meteorology ; Barrier: Not Applicable NUMARC Generic Basis: Valid means that a radiation montor reading has been confirmed by the operators to be correct. The 1000 mrsm <EDE + CEDE > and 5000 mrem <CDE> thymid integrated doses are based on the EPA l protective action guidance which indicates that public protectue actions are indcated if the dose exceeds 1 rem
<DE + CEDE > or 5 rem <CDE> thymid. This is consistent with the emergency class hva- ni for a Geners! l Emergency. This level consttutes the upper level of the desirable gradient for the Site <E>mergency. Actual !
meteorology is specifically identified in the initiating condition since it gives the most accurate dose assessment. Actual meteorology (including forecasts) should be used whenever possible. l I Integrated. doses are generaly not monitored in real-time. In establishing the emergency action levels, it is l suggested that a duration of one hour be assumed, and that the Eats be based on site boundary doses for either j
<EDE + CEDE > or <CDE> thymid, whichever is more limiting (depending on source term assumpoons). If induidual i site anahses indicate a longer or shorter duration for the period in which the substartial portion of the actwty l is released, these dose rates should be a4usted. l l
The FSAR source terms applicable to each monitored pathway should be used b conjuncton with annual average j meteorology in determining indications for the monitors on that pathway Plant-Specfc Information 10 CFR Part 20 was revised following the development of the NUMARC methodology. Calvert Cliffs uses the new rule as its basis for determining dose. EDE is the Effective Dose Equivalent as defined 510CFR20.1003. &DE is the Commtted Effective Dose Equivalent as defined in 10CFR20.1003. COEis the Commazed Dose Equivalent as defined in 1DCFR20.1003. Calvert Chffs does not have eitiver a penmeter redeuon monitonng system or automated reakime dose i assessment capability. Thus, the generic EALs recommended by NUMARC do not apply to the Calvert Diffs l Nuclear Power Plant. A description of the applicable monitorr and the methods used to calculate EAL values for the WRNGM is shown in RU1 Unplanned Radioacthe Release Exceeding 2 XTech Spec Lamcs for GREATER THAN 60 Minutes. Values for this IC are based on tne values shown in RU1 scaled up from O.114 mrum in an hour (le., houriy rate resulting in 2 X 500 mrem in one year) to 1000 (EDE + CEDE) mrem in an hour (10 m9v/h). . EALs 1 and 2 only apply if dose assessment capability is not available, i.e., wittout dose assessment. The preferred method of declaration is via EAL 3, with EALs 1 and 2 as backup methods, f required. Calvert Cliffs EAL Basis Document R22 June 15.1993 J
~~~ - - ~ , - - __.__._:~~
~~RADIDACTMTY FELEASE RG1 Threshold for RG5415 Scale up fmm RU1 uncorrected release mte of 3.6 E+5 pCi/see RG1 Value .1_000 mrem /h x 3.6 E+5 pCi/sec O.114 mmm/ hour (or .00114 mSv/h)~~
~~- 32 E+9 pCi/sec Read as 3 E+9 pC(/sec This value cormsponds to a conmntraton of about 50 CVcc and falls wthan the range of the WRNGM Thus. EAL 1 is wntten as:~~
Valid WRNGM (RG5415) Reading of AT LEAST 3 E+S Ci/sec for GREATER THAN 15 Minutes (Wchout Dose Assessment) The Main Steam Effluent Radiation Monitor System is described under C RS1, Off-Sce Dose of AT LEAST O.1 Rem (EDE + CEDE) OR O.5 Rem CDE Thymid. The appmpriate EAL value for this C was determined by scaling up the RS1 reamg to correspond to 1,000 mrern in one hour (10 m9v/h). RG1 Threshold for RL5421. Rh5422 RG1 Value - RS1 Value x 1000 mrem /h 100 mrem /h
~~- 02 rern/h x 10 - 2 rem /h (20 mSv/h)~~
The 2 rem /h (20 mSv/h) on 1/2-RL5421/5422 is based on assuming a single stuck open safetyvah. A value of 3 rem /h (30 m9vh) corTesponds to assuming a single stuck open atmospheric dump valve. Thus, EAL 2 m wntran as: 1 Valid Main 3 team Effluent Monitor (Rh5421. Rh5422) Reading of AT LEAST 2 rem /h for GREATER THAN 15 Mintus (Wehout Dose Assessment) 5 Valid rneans that the indcation is from instrumentation determined to be operable in accordance with the Technical Specifcations or has been venfed by other independent methods such as indications displayed on the control panels, reports fmm plant personnet, or radiological survey results. Based on the Mamh 14,1993 SG tube rupture event at Palo Verde Unit 2, the main steam effluent monitors (RL5421 Rh5422) may read N'8 immediatett following SG tube rupture and prior to reactor trip. However, given the short half-life of N, this should clear within the first minute following reactor trip. Dose consequences can be determined by use of ERPIP B22. Initial Dose Assessment Manual Calculation Methods, or by use of ERP!P B23, Dose Assessment Computar. Dose assessment will be performed in accordance with the new 10 CFR 20 scheduled to take effect January 1,1994. (. Calvert Chffs EAL Basis Document R23 June 15,1993 n
RADCACTMTY ELEASE Thus. EAL 3 is wntten t.s: Dose Assessment Determines integrated Accdent Forecast Dose OffGite is AT LEAST 1 rem (EDE + CEDE) Or 5 rem CDE Thyroid This corTesponds to doses of 1D rr& (EDE+ CEDE) and SD m9v CDE Thyroid, respectkely. j Sourt:e Documents / References /Calculatio.ns:
~~1. System Descriptions~~
~~No.15. Radiation Monitoring System~~
~~2. Emergency Response Ran implementation Procedoms~~
ERRPa1D. Main Steam RadioactMty Release Estmate

ERPIPB21. Accdental Radioaczkity Release Monitoring and Sampling Methods

ERPIP-822. Initial Dose Assessment Manual Calculation Methods

ERPIP-823. Dose Assessment Computer %
3. Radioactuity Release Emergency Action Levels J.B. Mcikaine JS8 Associates. Inc., September 199D

4. BG&E Intemal Memorandum. J. R. Hill (Nuclear Rant Operations) to R. L Wenderiich. CE Operations Subcommittae Meeting - Trip Report. April 16,1993

5. 1D CFR 2D Standards for Protection Against Radiation; F:nal Rule. 56 FR 2336D. May 21,1991 O
O Calvert Clrffs EAL Basis Document R24 June 15,1993
FISSION BARRIER REFERENCETAE INDICATORS FUEL CLAD BARRIER RCS BARRIER Loss-Not Applicaole loss EOP 8. Funcuanel Recom Pmeedum, e implemer from EOA-8. Steam Generecor Tube Rupture SAFETY FuNCiloN PotenvalLoss PotentialLoss STATUS / FUNCTIONAL b EOR 8. Functional Recovery Procedure. Can NOT Meet " Uncontrolled RCS Cooldown AND RCS Pmasure-C m sna RCS Heat Removal Acceptance Cntens AND Temperatum in the Nor>Opsmung Ares [Left of the RECOVERY Shutoown Cochng a NOT in Senace Cocidown Curve) OR EOR 8. Functional Recovery Procedure. Can NOT M Core and RCS Heat Removal Acceptance Cntene AN Shutdown Cochng is NOTin Sarwce Loss Loss Vehd Core Est Thennoccuois Readinge GREATE.R THAN RCS Succookng Can NOT Be Maintained AT LEAST 2$ TEMPERATURE 1mF PotencalLoss PotentialLoss- Not Applicable vaka Core Est Themocouole Readinge indcate Sucomeet Loss Loss Dose Rata et One Foot Fmm PASS Sample cf AT LEAST 40 Vehd RL5317A/B Reading of AT MAST 5 mm/h h mmm/h Hours AfterReacrorShut.down OR Vehd Rb5317A/B Reedog of AT MAST 1500 rem /h PotentialLoss- Not Applicable RADIATICN Wt.hin 2 Hours Aftar Reactor Shutdown OR AT LEAST 5% Fuel Clad Damage se Determined Fmm Core Camage Assessment J k#M7 PotentialLoss-Not Applicable Loss- Not Applicaole l
~~'Not Applicablel COOLANT LEAKAGE '- -~~
~~Potentialloss RCS Leenage Exceeds Available CVCS Capacity OR~~
~~~ ~_@' . ~ .^ - >~~
p
~~'m..~~
~~EDFL5. Loss of Cocient Accedent, orEORE. Function;~~
~~. Recovery Pmoedure, a knoiemented for RCS Lasked E^#~~
~~.:p:.~~
49
~
JNot Applicable ! : Not Applicable? % PRESSURE -' ir Loss-Not Applicable /
~~.. . .. 7, ;~~
WATER LEVEL (Not Appiicable? , ; ' Potentialloss .. Vshd RVLM3 Aeedeo et 29 inches and Tmndino Towertt O O SEC JUDGEMENT Any Concuan Which in the SECe Judgement indicates Loss or Any Conditson Which in the SECS Judoement Indicates Lost Potenuai Loss of the Fuel Ciad Bemer Based ort Potanuel Loss of the RCS Bemec Be6ed ort e immetent Bemer Degracecon Due to Sefsty Systam e knminentBemerDegradeuonDuetoSefstySystem , Performance Performance e Cecmoed Andrtv to Moner Bamere e Decreded Abihtv to Moner Romere I 1
~~
LE JUNE 15,1993 THREE BARRIERS AretustD q . CNTMT BARRER L P L P L P S, ,,,i cm - nt . m Loss-Not Apolicable n ; CLAD RCS CNTMT t -- L' tad - 1 Potentia' Loss '3 i , , ;' c ' - () 1
\
t () . EOPA Functional Recovery Procedure, e implemented AND Containment Environment Acceptance Cntana Con [ E NUT Be Me 3/3 =t V Loss- Not ApplicaDie LOSS OFAT YES GENERAL
~~F ' ' ^~~
PotentialLoss BARAG S7 EMERGENCY Veiid Core E;at Thennocouple Readings GREATER THAN i 13OO'F AND INCAEADING Loss t w2 EDM Loss of Coolent Accident, or EOPa Funczional Recowwy Procedure, e impemented AND Redemn L'a* Extemel to CNTMT Con NOT Meet Acceptance Cntans TWO BARRIERS ArttuitD . PotentialLoss Vered 5317A/B Reading of AT LEAST 14DCD rerr/h . Within 2 Houns After Reactor Shutdown ' on MD RCS %'M EDM Loss of Coolant Accedent, e impiamented AND LOCA e NOT Occumng Wehin CNTMT se bdicated by Aux I Buildma Sumo Alarms or Aux Buildino AMS Alerme v loss yu v L ( Lamicsge Pathway E;osts From inside CNTMT to Outside CNmr $Q 2/3 oR EMERGENCY L SGT@e Ruptune in Progrese AND Both of the Following- i e Afected SG Level Con NOT Be Meetsined LESS THAN + 50 t inches AND ! si e Affected SG Pressure GFEATER THAN 900 PSG ,__________________________j i PotentialLoss-Not 400ficable , f L 88 $oE g, ONLY ONE BARRIER Atreultu I g Rood unexoteined CPCMT Pressurs Decrease Following l inceimerases
~~ALERT. ;~~
~~I PotencialLoss =~~
~~'l > CNTMT Pressure of AT LEAST 50 PSIG And inerseeing o'~ . oR o~~
CNWT H2 Concentreten of ATLEAST4.0% w JihtiM;M Wae# Web - '* ffp ..,fQ 7 % sjy [ f 'QQm. ..u4?d ?Q N-)M L P L P L P g dgeo pdA .m., x . a
~~} . 8r "n'W["'pMp' - a igg ~g ;y3 liNdek .~~
w% f* CLAD RCS CNTMT C % e D '~ 4 or Any Conocion Whicn in the SEC's Judgement inchcetas Lose or Potental Lose of the CNTMT Bemer Beesd on:
. mninent Bemer oogr.demn oue to Sefey system l-per4annonce .
UNUSUAL I e Degruded Abiliev to Montor Bemere MQ 5U07280/(s6 - 0 ._ . _
O O FISSION PRODUCT BARRIER DEGRADATION I i
)
O l l
FISSION PRODUCT BARRER DEGRAEATION Erne _rgency Classification Levet UNUSUAL EVENT AppJicable Operational Mode _g: 1,2.3.4 Cabert Cliffs Initiating Condtiori: BU1 Loss OR Potentaalloss of CNTMT Barrier NUMARC Initiating Condtion: FU1 ANY Loss or ANY Potential Loss of Containment Barrier: Containment This C is entemd based on the Rssion Barrier Refemnce Table EAls discussed below. 1 i i l i
)
~~I i 1 l~~
\
m U 1 Cabert Cirffs EAL Basis Document B:1 June 15.1993 j
RSSION PRODUCT BARRER DEGRADATION E_menge_n_cy Classification Level: UNUSUAL EVENT Applicable Operational Modes: ALL
)
V Qalvert Cfdfs indstino Conditio_n: BU2 RCS Leakage NUMARC Recoonbon Categon. System Malfunction RUMARC Initiatino Conddon: SUS RCS Laakage Bamer: RCS NUMARC Generic Basis: This IC is included as an Unusual Event because it may be a precursor of more serious conditions and, as a msult, is considered to be a potential degradation of the level of safety of the plant. The 10 43PM> value for unidentified or pressure boundary leakage was selected because it is observable with normal control mom indications. Lesser values must generally be determined thmugh timeconsuming surveillance tests (e.g., mass balances). The EAL for identified leakage is set at a higher value due to the lesser significance of identified leakage in comparison to unidentified or pressum boundary leakage. In either case, escalation of this IC to the Alert level is via Fission Pmduct Barrier Degradation <EALs or the IC QA2, inability to Maintain Plant in Cold Shutdown >. Only operational modes in which fuelis in the mactor coolant system and the system is pressurized are specified. l Plant-Specific Information: , Technical Specification Section 3.4.62 specifes allowable RCS leakage as:
a. No Pressure Boundary Leakage i

b. 1 GPM unidentified leakage

c. 1 GPM total primary-to-secondary leakage through both steam generators and 0.1 GPM thmugh any one steam generator

d. 10 GCtA identified leakage fmm the RCS STPO27-1/2 is the daiy surveillance test procedure that the operators use to measum the amount of RCS leakage. The RCS GROSS leakage rato is based upon the following parameters: T,, RC Makeup Integrator reading, Boric Acid integrator reading, Diversion integrator reading, volume control tank level, pressurizer j pressure and pressurizer level, If the RCS GROSS leakage rate is calculated to be GREATER THAN 11D GPM, 1 ADNA is implemented. This 11 GPM threshold cormsponds to the net RCS makeup fmm one charging pump j in the normal CVt S alignment (44 GPM charging flow minus the total flow from reactor coolant pump sealleakoff and minimum Wwl. If the RCS GROSS leakage rate is calculated between 1D GPM and 11D GPM, the l difference between RCS GROSS leakage rate, Reactor Coolant Drain Tank (RCDT) inleakage, safety injection tank (SIT) outleakage Quench Tank (OT) inleakage, and the calculated SG leakage fmm CP@24. If the difference is GREATER THAN 1D GPM, the CRS is notified and with his approval, ADPGA is implemented.
' I CaNert Cirffs EALs have been written to be consistent with procedural mquirements. These leakage rates are w very similar to the NUMARC generic leakage. AOPEA specifies certain flow paths that can be isolated to ]
terminate RCS leakage. If isolation of the leakage path is successful (e.g., isolating a leaking pmssurizer power Cakert Cliffs EAL Basis Document B-2 June 15,1993
RSSION PRODUCT BARRER DEGRAEMTION operated relief va%). reactor operation can continue and thin EAL does not appy. However,if RCS leakage could not be isolated, then under these conditions the mactor wou'd have to be shut down in accordance with tachnical n specircations. The EAL language was pickad to assure that (1) leakage is greater than net RCS make-up flow threshold of 11 GPM, ar.d (2) Such laskage could not be isolated in accordance with procedurel requirements. Thus, the Cawrt Cirffs EAL is wntten as: 1 l ADF42A Excessive Raactor Ccolant Laakage. Is implementad For RCS Laskage Exceeding the Capacity of l One Charging Pump AND Reactor Shutdown is Required ) NUREG 1449 reises concams regarding events invoMng leakage through RCS temporary boundaries. RCS ' leakage EALs appy to all operational modes at Calvert Cirffs. This will assure that leakage is appropriateY addressed for cold shutdown and rewling modes and address NRC concems about leakage through temporary I RCS boundaries as they appy to EALs. l l Source Documents / References / Calculations: l l
~~1. Technical Specifications J~~
~~T5 3.4.62. Reactor Coolant System Leakage i 1~~
~~2. Abnormat Operating Procedures 1~~
~~ADF2A Excessive Reactor Coolant Laskage~~
3. Surveillance Test Procedure (STP) 027-1/2 RCS Leakage Evaluation

4. NUREG d % 9. Shutdown and Low 4)ower Operation at Commercial Nuclear Power Plants in the United States.
Draft for Umment, February 1992 P Cahrt Cirtfs EAL Basis Document 8:3 June 15,1993
i RSSION PRODUCT BARRER DEGRADATION Emesency Cassification Level: UNLEUAL EVENT ! [ Apsicable Doerational Modes: ALL kJ CaNert Cirffs intiatino Condtion: BU3 R2al Clad Degradation NUMARC Recoonition Cateoorv System Malfunction b NUMARC Initiatino Condition: SU4 Fuel Qad Degradation Barrier: Fuel Clad NUMARC Generic Basis: This IC is included as an Unusual Event because it is considered to be a potentia! degradation in the level of safety of the plant and a potential precursor of more serious pmblems. < Generic > EAL 1 addresses (Sce-Specirc) radiation monitor madings such as failed fuel monitors, etc., that provide indcation of fuel clad integrity.
~~< Generic? EAL 2 addresses coolant samples exceeding coolant technical specifcations for iodine spike. Escalation of this IC to the Alert level is via the < Fission Pmduct Barrier Degradaten EALs>~~
Plantepecific Information: A significant rise in the count rete on the Act.svey Montor or valid actuaton of the " RADIATION MONITOR LEVEL A Hi" alarm can be due to either fuel clad failure or to crud burst. In accortlance with AOPEA, the response to high RCS activity level is to noufy Plant Chemistry to perform a sample analysis to determine what radionuclides caused the radiation alarm. This means that the monitor indications are not suffeient alone to determine whether fuel clad damage has occurmd at Cabert Chffs. Thus, < Generic > EAL 1 is not appmpriate for use st Calvert Cliffs. Cad damage is determined from specific acuvey levels contained in reactor coolant samples. Per AOPEA, Fuel Rehabilty Plan Action Level 4 is defined as Dose Equivalent l 0' ' DEG) of at least the Technical Specification Section 3.4.8 limits. These are:
a. Not more than 1 Ci/ gram l DEG.

b. Not more than 1DD/5 Ci/ gram of gmss radioactivity.
The specific actMty of the reactor coolant may be as high as the limits defined by Technical Specifcation Figure j 3 4-1 for up to 48 hours. The lowest limt for this figure corresponds to SD uCi/grem l DEQ. Scafing down fmm the value shown for FCB3, Radiation, corresponding 15DO pCi/ gram l' ' DEO an RCS sample dose rate at one foot is computed as shown in the equation below.
]
I Calvert Orffs EAL Basis Document June 15,1993 ! B:4
FESION PRODUCT BARRER DEGRADATION RCS Samph Reading For 60 Wgram l* DEG O Refer to EAL FCB4, High RCS Activity BU3 Value = 601D_Zgram x 168 mrem /h = 6.7 mrem /h 1500 Ci/ gram Read as 6 mrem /h (D6 mSv/h) Thus, the EAL is wrtten as: Dose Rate at One Foot fmm RCS Sample of AT LEAST 6 mrem /h This cormsponds to a dose rate of 0.06 mSv/h. ce i Source Documents /Seferences/Calculadons: ,
~~1. Technical Specifcations~~
~~TS 34.8. Reactor Coolant System - Specific Activey~~
~~2. Abnormal Operating Proceduras~~
~~ADPEA Response to High RCS Activey~~
~~3. BG&E Fuel Degradation EALs Calculation Worisheet, JS8 Associates. February 18,1993 i~~
l l 1 l l Cahert Cl#fs EAL Basis Document June 15,1993 i B.5
FISSION PRODUCT. BARRER DEGRADATION Em_ergency Classification L.evet ALERT Applicable Doeretional Modes: 1,2,3,4 Cabert Cidfs initiatina Condition: BA1 Loss OR Potandal Loss of BTHER Fuel Clad BarTier OR RCS Banrier NUMARC Initiatina CondLJon: FA1 ANY loss or ANY Potential Loss of ETTHER Fuel Dad OR RCS Barrier. Fuel Clad. RCS This C is entered based on the Assion Barrier Reference Table EALs discussed be:ow. s l l l l l l Cakert Or!fs EAL Basis Document B.6 June 15,1993
r-FISSION PRODUCT BARRER DEGRADATION Emeroency Cassification Level: SITE EMERGENCY 6policable Doerational Moder: 1,2,3,4 Catvert CMfs initiatina Condch: BS1 Loss Or Patantial Loss of ANY Two Bamars NUMARC Initiatino Conos_r): FS1 Loss of BOTH Fuel Qad AND RCS DR Potential Loss of BOTH Fuel Clad AND RCS OR Potential Loss of EITHER Fuel Dad DR RCS AND Loss of ANY Addcional Barrier 13arrier- Fuel Clad. RCS. Containment Calvert Chffs logic es simphfied fmm the generic NUMARC logic based on the following considerations:
1. Human Factors - t is easier to understand and to remember the escalation from Alert to Site Emergency to General Emergency using the simpler logic.

2. Comprehensiveness - A comparison was made of the combinations of barvier losses and potential losses corresponding to See Emergency between the Calvert Cliffs logic and the NUMARC logic. This comparison is shown by Tables B-1 and B-2 below. All six NUMARC barrier loss / potential loss combinations (Table B-2) are add essed in the Calvert Chffs logic that addresses 12 combinations of barrier inss/ potential loss (Table B-1).
p 3. Escalation of SG Tube Ruoture Sequences-This logic change is consistent with NUMARC's intended scheme ( for classifying steam generator tube rupture sequences. No other sequences are signi5cantly affected by
' the Hssion Barrier EAL logic change. IC BU2. RCS Leakage, addresses smaller sized SG tube leakage that exceeds Tech Spec allowable but fall well within norTnal makeup capacity. SG tube leaks in this category should msutt in declaration of an Unusual Event. Fission Barrier EAL RCB4. Coolant Leakage, addresses tube breaks with leak rates that are somewhat larger than normal makeup capacity, but are readiy controlled in accordance with the EDPs. SG breaks of this category resut in declaration of an Alert due to potentialloss of the RCS Bamer. Larger speamm tube rupture events that can lead to SG overfill and prolonged releases off-sce are addressed by Fission Bamer EAL CNB4, Coolant Leakage. Leaks of this size resuh; in simutaneoush achieving the thresholds of RCB4 and CNB4 resut in declaration of a Site Emergency due to potential loss of the RCS barvier and loss of the CTMT barrier. Escalation to General Emergency would be based on further degradation of the RCS barrier and the subsequent potential loss of the Fuel Clad barrier.
This C is entered based on the Rasion Banier Reference Table EALs discussed below. \._ Catvert Cirffs EAL Basis Document June 15,1993 B.7
RSSION PRODUCT BARRER DEGRAIMTION S. O V Table B-1: SAE Barrier Loss / Potential Loss Combinations for CCNPP Logic Loss or Potential Loss of ANY Two Barriers Fuel Clad RCS Containment Potential Potential Potential Loss Loss Loss Loss Loss Loss
1. X X

2. X X

3. X X r%

4. X X
()
5. X X

6. X X

7. X X

8. X X

9. X X

10. X X

11. X X

12. X X 0
%) Catvert Clifs EAL Basis Document B.8 June 15.1993
RSSION PRODUCT BARRER DEGRAIMTON r l t i Table B-2: SAE Barrier Loss / Potential Loss Combinations for NUMARC Logic { Loss of BOTH Fuel Qad AND RCS OR Potential Loss of BOTH Fuel Qad AND RCS OR Potential Loss of EITHER Fuel DR RCS, AND Loss of ANY Addcional Banier Fuel Qad RCS Containment i Potential Potential Potential Loss Loss Loss Loss Loss Loss
1. X X

2. X X

5. X X

6. X X

7. X X 8.
~~9. 10.~~
~~11. X X 12.~~
Cat <ert Diffs EAL Basis Document B.9 June 15.1993
FESION PRODUCT BARRER DEGRADATION Emergency Classifestion Level: GEhERAL EMERGENCY Appicable Operational Modes: 1.2.3.4 Calvert Clrlfs initiating Condtion: BG1 loss of Two Baniers AND Potential Loss of Third Barrier NUMARC Initiatina Condtion: FG1 Loss of ANY Two Barriers AND Potential Loss of Third Bamer Barrier: Fuel Clad, RCS, Containment This C is entemd based on the Rssion Banier Refemnce Table EALs discussed below. O i Catert Cirffs EAL Basis Document B:10 June 15,1993
RSSON PRODUCT BARRER DEGRADATON NJ cm I FUEL CLAD BARRIER EALs O l 1 O : Ca%rt Cirffs EAL Basis Document B:11 June 15,1993
FISSION PRODUCT BARRER DEGRADATION I Ememen_cy_ Classification Level: PER FISSION BARRER REFERENCE TABLE apolicable Doerational Modes: 1.2.3.4 v Cakert Cliffs Ememency Action Level: FCB1 Safety Function Status /Runctional Recovery NUMARC Ememency Action Level: Fuel Clad 1 Critical Safety Function Status
Loss - Core Cooling-RED

Potentialloss- Core CoolingORANGE or Heat Sink-RED Barrier: Fuel Clad - The Fuel Clad Bamer is the < zirconium alloy > tubes that contain the fuel pellets.
NUMARC Generic Basis: This < Generic > EAL is for PWRs using Critical Safety Function Status Tree (CSFST) monitoring and functional recovery procedures. < > Plant-Specific information: Calvert Cl ffs does not use Critical Safety Function Status Tmes. Cabert Cliffs uses Safety Function Status Checks developed by the Combustion Engineering Dwners' Gmup (CE OG) which are based on logic similar to that used for CSFSTs developed for Westinghouse PWRs. However, there is no Safety Function Status Check condition that cormsponds directly to Core Cooling - RED path as a loss EAL Thus, the < Generic > loss EAL is incorporated into FCB2. This is better addressed by core exit temperature readings as shown in FCB2, Temperature, below. ' As stated above, the Potential Loss EAL should cormspond to loss of subcooling conditions and the Loss EAL should correspond to conditions indicating significant supert> eat. A review was made of the Emergency Operating Pmcedure basis information contained in CEN-152. Emergency Pmeedure Guidelines, to determine the applicable Calvert Chffs symptoms cormsponding to the generic conditions of interest,i.e., symptoms of inadequate core cooling sequences leading to com heatup and significant fuel clad damage. Satisfying the appropriate RCS and Com Heat Removal cnteria assures that adequate com cooling exists. Following a LOCA, there are two paths initialty available for RCS heat removal: heat transfer to the secondary side via the steam generators, and heat transfer via the fluid flowing out the break. Large break LDCAs have sufficient fluid flowing out the break to provide adequate heat removal. Small break LDCAs do not have sufficient fluid flowing out of the break to provide adeauste heat removal Therefore, under these conditions steam generstor heat mmoval is required in addition to break flow to assure that there is adequate core heat removal. For the largest breaks, the RCS depressurizes to an equilibrium pmssure with containment. In this condition, the RCS fluid is at a lower temperature than that of the steam generators. The steam generators. therefore, act as a heat source, supertisating any steam in the RCS which may be flowing thmugh the steams generators to the bmak. By cooling down the steam generators, heat input to the RCS is reduced. EOP-5. Loss of Coolant Accident, does not distinguish between large and small break LOCAs, and mquires steam generator heat removal be maintained at all times during a LDCA, if at all possible. Once RCS pressure and temperature are mduced. RCS heat removal can be provided by the Shutdown Cochng System (SDCS). Once the SDCS is placed in service, the steam generator heat sink capability is no longer necessary. In the event that the liquid inventory in the steam generators is not adequate to remove decay heat, a source of feedwater is unavailable, and the SDCS is not in service, the operator will transition to EOP-8 Functional Recovery Procedure. If RCS cooling through the steam generators cannot be restored, then the operator is ( instructed to implement once through RCS Cochng via the (pressurizer) PORVs before depleting the mmaining steam generator inventory. Calvert Chffs EAL Basis Document B:12 June 15.1993
FISSION PRODUCT BARRER DEGRADATION The applicable acceptance criteria for Core and RCS Heat Removal are shown on the Safety Function Status Checks. They are: O l Steam Generators Available for RCS Heat Remova!
~~1. Adequate secondary side liquid inventory in at least one steam generator as indicated by level between~~
~~-170 and +30 inches, and~~
2. Adequate source of feedwater available to assure continued liquid inventory available as indicated by Condensate Storage Tank level greater than 5 feet, and

3. Steam Generatom acting as effective heat sink as indicated by Cold Leg Temperatures (Ty constant or lowering.
~~Primary Side Conditions for Core and RCS Heat Removal~~
1. Adequate core heat removal as indicated by Core Exit Thermocouple readings less than superheated, and

2. Either of the following:
Natural cimulation established as indicated by the temperature difference between Hot Leg Temperature (TW) and Tm of between 10'F and 50'F or

Forced circulation effective as indicated by T,- Tm less than 10'F.
Based on the above discussion, the Potential Loss EAL is wri:2en as: EOP-8. Functional Recovery Procedure. Can NOT Meet Core and RCS Heat Removal Acceptance Criteria AND Shutdown Cooling is NOT in Service O Can NOTis used because the final safety function status is of concem, not merely the inability to meet certain , intermediate status check conditions. in service means that the SDCS is in the proper configuration for RCS heat removal (SDCS isolation valves open. LPSI pumps operating, etc.) and is considered " operable" as defined in the Calvert Cirtfs Technical Specifications Section 1.6. Source Documents / References / Calculations:
~~1. Emergency Operating Pmcedures~~
EDP-5. Loss of Coolant Accident

E0P-8. Functional Recovery Procedure
~~2. CEN-152. Emergency Pmcedure Guidelines I~~
i x l l Calvert Cliffs EAL Basis Document B:13 June 15.1993
RSSION PRODUCT BARRER DEGF%EMTION Calvert CIWfs Eme.rgency Action Level: f i FCB2 Temperatum NUMARC Ememency Action Leve_l-Fuel Qad 3 Core Ext Thermocouple Readings
~~Loss - GREATER THAN (SteGpecific) *F e Potential loss - GREATER THAN (SteSpecife) "F NUMARC Generic Basis:~~
The Loss EAL (SceSpecife) reading should correspond to signifcant supertsating of the coolant. This va'ue typical 9 corresponds to the temperature reading that indcates core cooling - RED in Fuel Qad Bartier EAL 1 , usually about 12OO'F. The Potential Loss EAL (Ste-Specirc) reading should correspond to loss of subcooling. This value typica!!y corTesponds to the temperature reading that indcates core cooling - ORANGE in Fuel Clad Barrier EAL 1, usually about 700 to 900'F. Plant-Specife information: Catvert Cliffs uses the generic value of 1200 *F as the fuel clad " loss' indicator. This is consistent with Attachment 3 of ERPIP BO2, Core Damage Assessment Using Com Exit Thermocouples. This shows that at Calvert Chffs, clad rupture due to high temperature is not expected for core exit thermocouple temperature readings of less than 1200 'F. Thus, the Loss EAL is wrcten as: Valid Corn Exit Thermocouple Readings GREATER THAN 1200 'F Vahd means that the thermocouple channel (s) are considered to be operable in accortlance with the Technical Specifcations. For consistency with Catvert Diffs EOPs, the Potential Loss EAL is written as: Valid Core Exit Thermocouple Readings indcate Supert> eat Source Documents / References / Calculations:
~~1. Emergency Response Ran implementation Pmcedures~~
~~ERDIN302, Core Damage Assessment Using Core Ext Thermocouples~~
~~2. Emergency Operating Procedures~~
~~EOF 5, Loss of Coolant Accident~~
~
~~EOPE, Functional Recovery Procedure~~
\ Catvert Orffs EAL Basis Document B:14 June 15,1993
FISSION PRODUCl" BARRER DEGRADATION Catvert Odfs Ememency Action Level: 1 FCB3 Radiation G NUMARC Ememengy_Agtion Levg Fuel Clad 2 Primary Coolant Acuvey Level
Loss - Coolant Activ'cy GREATER THAN [Sce Specirc) Value

Potendal Loss - Not Appicable Fuel Clad 5 Containment Radiation Monitoring

Loss - Containment Radiation Monitor Reading GREATER THAN (SceSpecifc) R/Hr

Pbtential Loss - Not Applicable NUMMC Sneric Basis:
(Fuel Qad 2) This (SceSpeedc) value corresponds to 300 pCi/cc 1,33equivalent. Assessment by the NUMARC EAL Task Fome indicates that this amount of coolant acrucy is well above that expected for iodine spikes and corTesponds to about 2% to 5% fuel clad damage. This amount of clad damage indcates significant clad heating and thus the Fuel Qad Barrier is considered lost. There is no equivalent Potential Loss EAL for this item. [ Fuel Clad 5) The (Site Specife) reading is a v alue which indicates release of reactor coolant, with an elevated activity ind'cative of fuel damage. into the containment. The reading should be calculated assuming the instantaneous release and dispersal of reactor coolant noble gas and iodine inventory associated with a concentration of 300 Ci/gm dose { y equivalent lia, into the containment atmosphere. Reactor coolant concentratsons of this magnitude are several times larger than the maximum concentrations (including iodine spiking) allowed within tachnical specifcations and are themfore indcative of fuel damage (appmximately 2 - 5% clad failure depending on core inventory and RCS volume.) This value is higher than that spected for RCS Barrier Loss <EAL 2>. Thus, this EAL indicates a loss of both the fuel clad barrier and the RCS barrier <and would indicate at least a Site Emergency classifcation>. There is no Potential Loss EAL associated wth this itam. Plant-S.pecific Information: The NUMARC Methodology is based on the assumption that the typical maximum iodine spike allowed in the Technical Spec # cations is about 60 Ci/cc DEG 1,33, and thus a value of 300 Ci/cc is readily distinguishable. For Cabrt Cliffs, the cormsponding iodine spike allowable value is 270 pCi/cc DEG I,33/ Technical Specifcation Figure 3.4.B-1. Given the uncertainties in the sample analysis process, the Calvert Cidfs allowable iodine spike value can be indistinguishable fmm the generic value. The site specife value was determined by calculating various coolant radenuclide concentrations postulated to resut from 5% gap release at Calvert Cliffs. This corresponds to about 1500 Ci/cc DEQ 1 ,Under33 those condcions, PASS sampling using 12.5 ml pmssurized bomb samples would be used. The cormsponding values for a PASS sample are:
42 mrem /h (0.42 mSv/h) at 1 foot due to radoiodines (unpressurized sample)

22 mrum/h (OD22 m9v/h) at 1 foot due to noble gases

44 mrem /h (0.44 m9s/h) at i foot for pressurized sample t'
i t Cabrt Cirffs EAL Basis Document June 15,1993 B:15
RSSON PRODUCT BARRER TGRADA'nON Themfore, whether the sample is pressurized or not makes little difference, as the largest contribution to the dose rate after 1 hour decay is from the radio-iodines. O Thus, Loss EAL 1 is wntran as: ( Dose Rate at One Foot fmm PASS Sample of AT LEAST 40 mrem /h This corresponds to a dose rate of 0.4 mSv/h. Per Reference 3,this also results in a dose rate at one foot fmm an unshielded RCS sample of about 168 mrem /h. (1.7 mSv/h) The plant specific containment monitor radiation V.ues were determined from ERRPEOi, assuming 5% fuel clad damage. This pmcedure can be used to detent,ine the containment radiation monitor readings resulting fmm 5% fuel clad failure using Attachment 2 and assuming no power correction. The radiation monitor reading (1-RL5317A & B,2-Rh5317 A & B) corvesponding to 5% fuel clad failure at 2 hours after shutdown is about 3,500 rem /h (35 Sv/h). Thus, Loss EAL 2 is wntzen as: Valid RF5317A/B Reading of AT LEAST 3,500 rem /h Within 2 Hours After Reactor Shutdown Valid means that the appleable radiation monitoring channel (s) are considemd to be operable in accordance with the Technical Specifcations. The EAL uses the value of 2 hours after the initiating event (assumed to closely corsespond to the time of reactor shutdown) for simplicity in presentation to the Shift Supervisor acting as the See Emergency Coordinator (SEC). The two hour point was also picked because it allows ample time for transfer of the SEC duties to outside the Control Room. Technical support personnel can s!so use ERPIFL801. -802,-803, and -804 to determine core damage. Thus, Loss EAl 2 is wntten as: AT LEAST 5% Fuel Clad Damage As Determined From Core Damage Assessment Source Documents / References / Calculations: l J
~~1. Technical Specifcations~~
~~Figum 3.4&1, Dose Equivalent I-131 Primary Coolant Specife Activity Limit Versus Percent of Rated Power With the Primary Coolant Specife Activity > 1.0 Ci/ Gram Dose Equivalent 5-131~~
~~2. Emergency Response Plan implementation Procedures~~
ERPIPBO1, Core Damage Assessment Using Containment Radiation Dose Rates {

ERP1P402, Core Damage Assessment Using Core Exit Thermocouples

ERPIFL803 Core Damage Assessment Using Hydrogen

ERPIPEO4, Core Damage Assessment Using Radiological Analysis of Samples
~~3. BG&E Fuel Degradation EALs Calculation Worksheet, JSB Associates February 18,1993 l 1~~
Calvert Cirffs EAL Basis Document June 15,1993 B:16
FISSION PRODUCT BARRIER DEGRADATDN Cabert Chffs Ememe.nc_y Action Level: FCB4 Reactor Vessel Water Level NUMARC Ememency Action Level: Fuel Dad 4 Reactor Vessel Water Level
Loss - Not Applicable

Potential loss - Level LESS Th N (Site 6pecific) Value Barrier: Fuel Clad h!UMARC Generic Basis:
~~There is no~~
Loss

EAL corresponding to this item because it is better covemd by the other Fuel Clad Barrier

Loss" Eats.
~~The (site-specific) value for the~~
Potential Loss

EAL corresponds to the top of the active fuel For sites using CSFSTs, the

Potential Loss

EAL is defined by the Core Cooling - ORANGE path. The (sitespecific) value in this EAL should be consistent with the CSFST value.
Plant-Soecific Information: As part of its inadequate Core Cooling instrumentation, Cabert Cliffs uses a reactor vessel level monitoring system (RVLMS) that is displayed to the operators and can measure water level from the top of the fuel alignment plate to the top of the reactor vessel head. The bottom of this instrument's span closely corresponds to the (Srte-Specific) water indication of Potential Loss used by NUMARC. At the bottom of the pressurizer, the RVLMS initiates the first alarm. From AOP-38, the transition to the EOPs is contingent upon the decrease of the pressurizer level. Per AOP38, Attachment 14, a 29' RVLMS indication cormsponds to the bottom of the hot leg elevation and is the sixth (6th) RVLMS alarm. The threshold corresponding to the bottom of the hot leg with a trend to zero thus is chosen because it is readiy mcognizable by operators (i.e., consistent declaration) and indicates a severe loss of inventory. Thus, the Potential Loss EAL is wrcten as: Valid RVLMS Reading at 29 inches and Trending Toward D Valid means that the applicable vessel level monitoring channel (s) are considemd to be operable in accordance with the Technical Specifications. Source Documents / References / Calculations:
~~1. Abnormal Operating Pmcedums~~
~~ADN38 Attachment 14. RCS Levels~~
~~2. Updated Final Safety Anahsis Report~~
~~Section 7.5.9, inadequate Core Cooling Instrumentation Cakert Cirffs EAL Basis Document B:17 June 15,1993~~
~~RSSION PRODUCT BARRER DEGRADATION Cakert Cltffs Eme_rgency Action Level:~~
~~/7 Q FCBS SEC Judgement NUMARC EmeIgency Action Level:~~
Fuel Dad 6 Other (SiteSpectfc) indications Fuel Dad 7 Eme pency Director Judgement NUMARC Generic Basis: [ Fuel Clad 6] This EAL is to cover other (site-specific) indications that may indicate loss or potential loss of the Fuel Dad barrier, ! including indications fmm containment air monitors or any other (sitespecific) instrumentation. (Fuel Dad 7] This EAL addresses any other factors that am to be used by the <SEC> in determining whether the Fuel Clad barrier is lost or potentially lost. In addition, the inabilty to monitor the bemer should also be incorporated in this EAL as a factor in <SEC> judgement that the barrier may be considered lost or potentially lost. (See IC <EG1, Prolonged Station Blackout > for addeional information.) Bant.@pecific Information: Other indications were also considered and no additional reliable indications of fuel clad loss or potential loss could be determined. Thus, the generic "Other (SiteSpecrSc) Indications
~~Fuel Clad EAL does not apply to the Calvert Cirffs Plant.~~
Per the Emergency Response Plan, the Ste Emergency Coordinator (SEC) is the title for the emergency dimetor function at Calvert Cliffs. SEC considerations for determining whether any barrier Loss or Potential Loss include imminent degradation, bamier monitanng capabilty, and dominant accident sequences. Andcipated degradation of ANY Bamer within 2 hour based on a projection of curTent safety system performance is considemd to be imminent Barrier degradation. This must be considered by the SEC for timely declaration of a General Emergency. The term imminent refers to the inability to reach final safety acceptance before completing all checks. Decreased barrier monitonog abilty fmm loss of/ lack of reliable indicators must also be considered by the SEC when judging whether a Barrier may be Lost or Potentialty Lost. This assessment should also include instrumentation operabilty concems, medings from portable instrumentation, and consideration off+ite monitoring msults. Dominant accident sequences will lead to degradation of all Barriers. Such sequences can lead to entry into EOP-8 Functional Recovery Procedure. The SEC should also consult Station Blackout and ATWS ICs. as appropriate, to assure timely emergency classifcation declaration. i Thus, the EAL is wntran as: Condrbons Which in the SEC's Judgement Indicata Loss or Potentia! Loss of the Fuel Dad Bamer Based on:
Imminent Barrier Degradation Due to Safety System Performance

Degraded Abilty to Montor Barrier Source Documents / References / Calculations: ;
~~1. Emergency Response Plan Calvert Cliffs EAL Basis Document B:18 June 15.1993~~
RSSION PRODUCT BARRER DEGRADATON w. RCS BARRIER EALs O Calert Cirffs EAL Basis Ebcument B:19 June 15.1993
RSSION PRODUCT BARRER DEGRADATION Ememency Dassification Levet FER FISSION BARRIER REFERENT TABLE Apolicable Doerational Modes: 1,2,3,4 v Cabert Drffs_g_mgrgency Action Levet RCB1 Safety Function Status / Functional Reccuery NUMARC Ememen_cy Action Level: RCS 1 Crtcal Safety Function Status
Loss - Not Appleable

Pbtentialloss - RCS IntegrityREO OR Heat Sink-RED RCS 5 Other (SMSpecific) ind' cations parrier: RCS - The RCS Barrier includes the RCS primary side and its connections up to and including the pressurizer safety and relief vakes, and other connections up to and including the primary isolation valves.
NUMARC Generic Basis: (RCS1] This EAL is for FWRs using Creical Safety Funaion Status Tme (CSFST) monitoring and functional recovery procedures. < > There is no Loss F#1. associated with this item. [RCS 5] This EAL is to cover other (siteSpecifc) indcations that may indcate loss or potential loss of the RCS barrier, p including indcations from containment air monitors or any other (siteSpecife) instrumentation. Q Plant-Specife information: 1 Calvert Cirffs does not use Crecal Safety Function Status Trees. Cabert Drffs uses Safety Function Status Checks developed by the Combustion Engineering Owners' Group (CE OG) which are based on logic similar to that used for CSFSTs developed for Westinghouse PWRs A mview of plant design information and pmcedums showed that an appropriate sita-specife Loss EAL could be developed based on an EDP transition. Contingency actions related to maintaining RCS Cooling, RCS Inventory, and RCS Pressure functions in accordance with EDP6 dime the operators to EDPB, Functional Recovery Pmcedure, i Thus, the Loss EAL is wntten as: EDPa, Functional Reccrsery Procedure, is implemented fmm EDP6. Etsam Tube Rupture 1 The generic indications of concem for the Potential Loss EAL corTespond to exceeding Pressurized Thermal Shock j (PTS) cooldown limits or the deterinination of loss of secondary heat sink. These are discussed below. Among the Safety Functens to be maintained is RCS Pmssure Contrut The purpose of maintaining RCS Pressure j Contml is to maintain the RCS inventory in a subcooled condtion to provde an adequate cooling rymdium for the I core, and to prwent the loss of inventory out of a reref vatve. Per EDP4, the potential exists for pressurized thermal shock fmm an excessive cooldown rete followed by a repressurization. l The EDPs require that the plant condtions be maintained within the limits of the RCS Pmssure-lemperature
~~\ Curve fwhich is shown as A::achment 1 to the EDPs). Uncontrolled RCS cooldowns which result in pmssure-~~
[N temperatum condtions to the left of these curves (based on the combinations of Reador Coolant Pumps in I Catsert Drffs EAL Basis Document B20 June 15,1993
)
I
RSSION PRODUCT BARRER DEGRAt% TION operation), which'is labeled as the Non-Operating Area, is the condition which most closely cormsponds to the NUMARC concem. Thus, based on the above Potential Loss EAL 1 is wrtten as: Uncontmiled RCS Cooldown AND RCS Pmssure-Temperatum in the Nonoperating Ama (Left of the Cooldown Curvel Uncontm/ led means that the RCS cooldown was not the result of deliberate action performed in accordance with plant procedums and exceeds sliowable vessel cooldown limits. The applicable acceptance criteria for Com and RCS Haat Removal via the Steam Generators are discussed under FCB1 above. Once RCS pressum and temperature are reduced. RCS heat removal can be primded by the Shutdown Cooling System (SDCS). Once the SDCS is placed in service, the steam generator heat sink capability is no longer necessary. On the basis of the above discussion, Potential Loss EAL 2 is wruen as: EOPB Functional Recmery Procedure. Can NOT Meet Com and RCS Heat Removal Acceptance Crteria AND Shutdown Cooling is NOT in Service Can NOT is used because the final safety function status is of concem, not memty the inabilty to meet certain interinediate status check conditions. In service means that the SDOS is in the proper configuration for RCS heat removal (SDCS isolation vatves open, LPSI pumps operating, etc.) and is considered ' operable
~~as defined in the Calvert Cirffs Technical Specifications.~~
~~Source Documents /Refemnces/ Calculations:~~
~~1. Emergency Operating Pmcedures~~
EDPO. Post-Trip immediate Actions

EOP 1, Reactor Trip

EDP-3, Loss of Feedwater

EDP4. Excess Steam Demand Event

EDP-5, Loss of Coolant Accdent

EOPE, Steam Generator Tube Ruptum

EDPB, Functional Recovery Pmeedure
2. Emergency Operating Pmcedures Attachment 1

3. CEN-152, Emergency Procedure Guidelines l
1 I Calvert Cliffs EAL Basis Document B21 June 15,1993 l
RSSION PRODUCT BARRER DEGRt4ATION Cabert Chffs Ememency Action _(.evel: RCB2 Tamparature NUMARC Ememence Action Level: RCS 2 RCS Leak Rate e loss- GREATER THAN Available Makeup Capacty as indicated by a Loss of RCS Subcooling RCS 3 SG Tube Rupture e Loss -(Sce-Specific) <!ndicatiore NUMARC Generic Basis: (RCS El The
Loss

EAL addresses condtions where leakage from the RCS is greater than available inventory contml capacity such that a loss of subcooling has occurmd. The loss of subcooling is the fundamental indication that the inventory control systems are inadequate in maintaining RCS pressure and inventory against the mass loss thmugh the leak. <>
~~<* Potential Loss~~
~~EALs am addmssed in C RCB4, Coolant Laakage>~~
~~(RCS 3) This EAL is intended to address the full spectrum of Steam Generator (SG) tube rupture events in conjunction with < Loss EAL CNB4 and FCB EALs> The " Loss~~
EAL addresses ruptumd SG(s) with an unisolable Secondary Line Break corresponding to the loss of 2 of 3 fission pmduct barriers (RCS Barrier and Containment Barrier-this EAL will always result in <1.oss EAL CNB4>). This allows the direct miease of radioactrve fission and e activation products to the environment. Resultant off-site dose rates are a function of many variables. Examples f include: Coolant ActMty. Actual Leak Rate. SG Carry Over, lodine Partitioning. and Meteorology. Therefore, dose assessment in accordance with C <RG1, Off6te Dose of AT LEAST 1 Rem (EDE

CEDE) OR 5 Rem CDE Thymid>
is required when there is indication that the fuel matrix / clad is potentially lost. (Sita-specific) indication should be consistent wth the diagnostic actrvtaes of the Emergency Operating Procedures (EDPs), if available. This should include indication of reduction in primary coolant inventory, increased secondary radiation levels, and an uncontrolled or complete depressuristion of the ruptured SG. Secondary radiation incmases should be observed via radiation monitoring of Condenser Air Elector Discharge. SG Blowdown, Main Steam, ond/or SG Sampling System. Determination of the "uncontmiled* depressurization of the ruptured SG should be based on indication that the pressure decmase in the ruptured steam generator is not a function of operator action. This should prevent declaration based on a depmssurimtion that results from an EDPinduced cooldown of the RCS that does not involve the prolonged release of contaminated secondary coolant fmm the affected SG to the envimnment. This EAL should encompass staam breaks, feed breaks, and stuck open safety or relief vatves. Elant-Specific Information: A review of EDP6 shows that the minimum acceptable RCS subcooling value is 25 F. Following ADP2A < l procedures for Excessbe RCS Leakage Exceeds One Charging Pump in Modes 1 & 2, EDPO, Post Trip immediate Actions is intiated. After the completion of EDPO, EDP.6 requires that the RCS Subcooling be maintain a I minimum of at least 25 "F. Failing this criterion will pmmpt entry ireo EDP-5. In addtion. ADPEA pmcedums { for Excessive SG Tube Leakage with LTDP (i.e., plant initialty in Mode 3) require that the RCS Subcooling maintain l at least 25 'F. For consistency with pmcedural requirements the lower value of subcooling is used for this EAL l l i l i Calvert Cirffs EAL Basis Document 822 June 15.1993 ;
\
1
RSSON PRODUCT BARRER DEGRADATION ] Thus, the Loss EAL is wntren as: l RCS Subcooling Can NOT Be Maintained AT LEAST 25 'F L/ Sourte Documents / References / Calculations:
~~1. Abnormal Operating Procedures~~
~~AOPEA, Excessbe Reactor Coolant Leakage s~~
~~2. Emergency Operating Pmeedoms~~
EOPO, Post Trip immediate Actions '

EDP-1. Reactor Trip

EDP-5, Loss of Coolant Accident

EDP4. Staam Generator Tube Ruptum em t
O Cakert Cirffs EAL Basis Document B23 June 15,1993
FISSON PRODUCT BARRER DEGRADATION Catvert Chffs Emeroency Action Level: O %) RCBS Radiation NUMARC Emergency Acdon Level: RCS 4 Containment Radiation Montoring
Loss - Containment Rad Monitor Reading GREATER THAN (Sitespecific) R/Hr

Pbtential Loss - Not Applicable NUMARC Generic Basis:
The (Srte-specife) reading is a value which indicates the misase of reactor coolant into the containment. The reading should be calculated assuming the instantaneous release and dispersal of the reactor coolant noble gas and iodine inventory assocated wrth normal operating concentrations (i.e., within TS) into the containment atmosphem. The mading will be less than that specifed for Fuel Clad Barrier <EAL 2>. Thus, this EAL would be indicative of an RCS leak only. < >. However if the ste specific physical location of the containment radiation monitor is such that radetion from a cloud of released radiation gases could not be distinguished fmm radetion fmm adjacent piping and components containing elevated reactor coolant acchty, this EAL should be omitted and other site specific indications of RCS leakage substruted. There is no Potential Loss EAL associated with this tem. Plant-Specific information: Only small amounts of noble gases would be dissolved in the reactor coolant t~cause or very h'gh clad integrity. ( The EAL uses a value of 5 mm/h (50 mSv/h) for ease of identrfcation using RF5317 A & B mtnitors, because it in the first decade of the log scale and will be readable by the operators. Thus, the Loss EAL is wntten as: Valid RL5317A/B Reading of AT LEAST 5 R/h Wohin 2 Hours After Reactor Shutdcrwn i By specrfyi.1g the time of the reading as being after reactor shutdown, it also eliminates from consideration such factors as " shine' and f416 effects on the detectors Valid means that the applicable radation monitoring channel (s) are considemd to be operable in accordance wth the Technical Specifcations. The EAL uses the value of 2 hours after the initiating event (assumed to closefj correspond to the time of reactor shutdown) for simplicity in presentation to the Shft Supervisor acting as the See Emergency Coortlinator [SEC). l The two hour point was also picked because et allows ample time for the trensfer of the SEC duties to outside l the Control Room. l l Soume Documen_ts l References / Calculations:
~~1. Emergency Response Plan implementation Procedures~~
~~ERPlP-801. Core Damage Assessment Using Containment Radstion Dose Rates O l 4~~
Larvert Cliffs EAL Basis Document B24 June 15,1993 I
i RSSION PRODUCT BARRER TGRADATION Catvert Cliffs Ememency Action Level: / (% RCB4 Coo! ant Leakage NUMARC Ememency Action Levet RCS 2 RCS Leak Rate
Fbt.endalloss - Unisolable Leak Exceeding the Capacey of One Charging Pump in the Normal Charging Mode RCS 3 SG Tube Rupture o Potandalloss - (Sta-Specirc) indcation that a SG is Ruptured and Has a Nor>4solable Secondary Line Break OR (Ste Specife) Indcation that a SG is Ruptured and a Pmlonged Release of Secondary Coolant is Occumng Fmm the Affected SG to the Environment RCS 5 Other (Ste-Specrfc) Indications NUMARC Generic Basis:
~~[RCS 2. RCS 3)~~
~~<t.oss EALs am addmssed under C RCB2, Temperature > <>~~
The Potential Loss EAL is based on the inabilty to maintain normal liquid inventory wthin the Reactor Coolant System (RCS) by normat operation of the Chemical and Volume Control System which is considered as one centrifugal charging pump discharging to the charging header. <This indcation, apptying to any RCS leakage including pnmarytCHinecondary leakage > assures that any event that results in signifcant RCS inventory shrinkage or loss (e.g., events leading to reactor scram and ECCS actuation) will result in no lower than an
Alert *
[ ( emergency classification. [RCS 5) This EAL is to cover other (ste-specific) indications that may indicate loss or potential loss of the RCS barrier, including indications fmm containment air monitors or any other (site specife) instrumentation. Plant Spftcife information: The Calvert Chffs Chemical and Volume Control System (CVCS) uses three positive displacement horizontal pumps with a capacity of 44 GPM each. The pressurizer level control program reguletas ietdown punfication subsystem flow by adjusting the letdown flow control vatve so that the reactor coolant pump (RCP) controlled leak-off plus the letdown flow matches the input from the operating charging pump. Equilibrium pmssurizer level conditions may be disturbed due to RCS temperature changes, power changes, or RCS inventory loss due to leakage. A decrease in pressurizer water level below the programmed level will result in a cortrol signal to ctart one or both standby charging pumps to restore water levet. An increase in pressurizer water level above the programmed level will result in a control signal to increase letdown punfcation flow rete and initiate a backup signal to stop the two standby charging pumps. A start signal is sent to all three charging pumps on a Safety injection Actuation Signal (SlAS), aligning the charging pumps suction to the Bonc Acid Storage Tanks (BASTS) via the bonc acid pumps. All three charging pumps will then inject highly concentrated boric acid into the RCS to ensure that the reactor is shutdown. Potential Loss of the RCS cormsponds to condtions where the CVCS can not maintain pressunzer water level wthin normal limes requiring transition into the EOPs when the reactor is initialty crccal. Thus Potential Loss EAL 1 is wntten as: RCS Leakage Exceeds Available CVCS Capacity Calvert Cirffs EAL Basis Document B25 June 15,1993
FISSION PRODUCT BARRER EGRADATION However, revew showed that an appropriate ste specife Potential Loss EAL could be developed based on entry into EOP-5, Loss of Coolant Accident or EON 3. Functional Recovery Pmeedure, for an RCS leak. r^g Thus, Potential Loss EAL 2 is wn: ten as: EOP-5, Loss of Coolant Accdent, Or EDPE, Fun conal Recovery Pmcedure, is implemented for RCS Laakage Soume Documents / References / Calculations:
~~1. Abnormal Operating Procedures~~
~~ADN2A, Excessive Reactor Coolant Leakage~~
~~2. Emergency Opersting Pmeedums~~
EDP-5, Loss of Coolant Accident

EOP4, Functional Recovery Pmcedum
3. Surveillance Test Pmcedure (STP) 027-1/2. RCS Leakage Evaluation

4. Updated Rnal Safety Analysis Report
~~Section 9.1, Chemical and Volume Contml System r~~
O Calvert Cirtfs EAL Basis Document 826 June 15,1993
RSSION PRODUCT BARRIER DEGRADATION Cahrert Cliffs Ememency Action Level: RCBS SEC Judgement NUMARC Ememency Action Level: RCS B Emergency Dimaor Judgement NUMARC Generic Basis: This EAL addresses any other factors that are to be used by the <SEC> in determining whether the RCS barrier is lost or potentially lost. In addtion, the inabilty to monitor the barvier should also be incorporated in this EAL as a faaor in <SEC= judgement that the barrier may be considered lost or potentially lost. [See also <3C EG1, Prolonged Station Blackout >, for addtional information.) Plant-Sp_ecific Information: Per the Emergency Response Plan, the See Emergency Coordinator (SEC) is the title for the emergency director function at Cahrert Chffs. SEC considerations for determining whether any barrier Loss or Potential Loss include imminent degradation, bamier monstoning capability, and domenant accident sequences. This information is included on the Assion Pmduct Barrier reference page which is to be revewed by the SEC befom using the Fission Pmduct Barrier Table. Anticipated degradation of ANY Barrier within 2 hour based on a projection of curmnt safety system performance is considered to be imminent Barrier degradation. This must be considemd by the SEC for timely declaration of a General Emergency. The term imminent refers to the inabilty to reach final safety acceptance before completing all checks. Q Decreased barrier monitoring ability fmm loss of/ lack of mliable indcators must also be considemd by the SEC when judging whether a Barrier may be Lost or Potentially Lost. This assessment should also include instrumentation operabilty concems, madings fmm portable instrumentation, and consideration off-site monitoring i results. l l Dominant accident sequences will lead to degradation of all Barriers. The SEC should also consult Station ! Blackout and ATWS ICs, as appropriate, to assure timely emergency classif' cation declaration. Thus, the EAL is wrcen as: Condtions Which in the SEC's Judgement Indicate Loss or Potential Loss of the RCS Barrier Based on: e imminent Barrier Degradation Due to Safety System Performance
~~Degraded Abilty to Monitor Barrier Sourse Documents / References /_ Calculations:~~
~~1. Emergency Response Plan I l~~
~~Cahert Cirffs EAL Basis Document B:27 June 1S,1993~~
~~. . .. ~ ____ _ _. _ _ . _~~
~~FISSON PRODUCT BARRER NGRADATION O L i i i CONTAINMENT BARRIER EALs O ;~~
~~.i O i i~~
i Cahert Orffs EAL Basis Document B28 June 15.1993
i RSSION PROOLCT BARRER TEGRADATION l E_memencyJQiassification Level PER FISSION BARRER REFERENCE TABLE Apolicable Operational Modes: 1,2,3.4 ( CaNert Diffs Ememency Action Level ) CNB1 Safety Funcuon Status /Funcbonal Recovery NUMARC Ememency Action Level: Containment 1 Cntcal Safety Funcion Status e loss- Not Appleable
Pbtential loss - Containment - Red Containment 7 Other (Site 6pecifc) Indications mm Barrier: CNTMT -The CNTMT Barrier includes the containment building, its connections up to and including the !
J outermost containment isolation valves. This banier also includes the main steam, feedwater, and blowdown line extensions outside the containment building up to and including the outermost secondary side isolation vake. NUMARC Generic Basis: [ Containment 1] This < Generic > EAL is for PWRs using Cntcal Safety Funcion Status Tree (CSFST) monitoring and functional recovery procedures. < >Thus, this EAL is primari9 a disenminator between Site <E>mergency and General Emergency representing a potentialloss of the third banier. Them is no ' Loss' EAL associated with this item. s (Containment 7] This <Genenc> EAL should cover other (site-specific) indications that may unambiguously indicate loss or potential loss of the containment barrier, including indications fmm area or ventilation monitors in containment annulus or other contiguous buildings. If site emergency operations procedums provde for venting of the containment during an emergency as a means of pmventing catastrophic failum, a Loss EAL should be included for the containment barrier. This < Generic > EAL should be declamd as soon as such venting is imminent. Containment venting as part of mcovery actions is classified in accordance with < Radiation Releases ICs>. BanJ-Specific Information: Calvert Orffs does not use Cntcal Safety Function Status Tmes. Them is no direct equivalent to the generic containment status tme ' potential loss' EAL at Cakert Orffs. Calvert Orffs does not include containment venting as part of its EDPs. Other conditions of intemst are almady addressed by the other CNTMT EALs below. Therefore, there is no Loss EAL in this category at Cabert Diffs. The potential loss addresses the inabilty to maintain mquired containment condcions folk.ving entry into the , I Emergency Operating Pmeedures. This corresponds to a potential challenge to containment intagnty and thus is an appropriate potential loss EAL l i Thus, the Potential Loss EAL is wntren as: EDP-B. Funaional Recovery Procedure, is Entered AND Containment Environment Acceptance Cnteria Can NOT Be Met O l Calvert Diffs EAL Basis Document June 15,1993 l B29 l l
F1SS10N PRODUCT BARRER DEGRADATION Can NOTis used because the abilty to meet the final acceptance critaria is the appmpriata concem, not whether interTnediata acceptance criteria are not being achieved at any given moment.
,y
(
~~) Source Documents /Refemnces/ Calculations:~~
~~1. Emergency Operating Pmcedures~~
~~EOP-8. Functional Recovery Pmcedure I~~
C I i I 1 l l l l l i Calvert Cirffs EAL Basis Document gno June 15,1993 i l
FISSION PRODUCT BARRIER DEGRADATION Erne _rgency Classification Level: PER FISSION BARRER REFERENCE TABLE applicable Operational Modes: 1,2,3,4 Calvert Chffs Emergency Action Level CNB2 Temperature NUMARC Ememency Action Level: Containment 6 Core Exit TArmocouple Readings
Loss- Not Applicable

Potential Loss - Core Exit Thermocouples in Excess of 12007 and Restoration NOT Effective Within 15 Minutes: OR Core Exit Thermocouples in Excess of 700'F Web Reactor Vessel Level Below Top of Active Fuel and Restoration Pmcedures NOT Effecdve Within 15 Minutes NUMARC Generic Basis:
in this EAL the function restoration p ocedures are those emergency operating pmcedures that address the recovery of the com cooling critical safety fundions. The procedure is considered effective if the temperature is decreasing or if the vessel water level is increasing. The conditions in this potential loss EAL repmsent < imminent > melt sequence which, if not corvected, could lead to vessel failum and an increased potential for containment fsilure. In conjunction with the core exit thermocouple EALs in the Fuel and RCS barrier columns, this EAL would result in the declaration of a General Emergency-loss of two barriers and potential loss of the third. If the function restoration procedures are ineffective, there is no
~~' success~~
~~path.~~
Severe accident analyses (e p., NUREG-11501 have concluded that function restoration procedums can arrest com degradation within the reactor vessel in t'signifcant fraction of the core damage scenarios, and that the likelihood of containment failure is very small in these events. Given this,it is appropriate to provide a reasonable period to allow function restoration procedures to arrest the core melt sequence. Whether or not the procedures will be effective should be apparent within 15 minutes. The <SEC> should make the declaration as soon as it is determined that the procedures have been, or w;ll be, ineffective. The reactor vessel level chosen should be consistent with the emergency msponse guides applicable to the facility. There is no
~~Loss" EAL associated with this item.~~
Plant-Specrfic information: As described in EAL FCB4, Reactor Vessel Water Level, the RVLMS measures water level to slightly above the top of the active fuel Therefore, the generic condition of water level below the top of the active fuel and temperatum grearar than 700 F does not apply to Calvert Cirffs. EOP 8, Funaional Recovery Pmcedum, would be entered on symptoms of inadequate core cooling. This includes core exit thermocouples reading superheat. The functional recovery procedure would be entered well before 1200 F core exit temperature is achieved, which is the threshold for clad rupture due to high temperature used in EAL FCB2. The clear intent of the NUMARC methodology is to pmvide a higher threshold for containment " potential loso* above that of fuel clad
loss

at 12OO F for the core damage sequences of concem, if core exit temperature i continued to increase above this value, it would cleariy indicate that functional recovery of RCS heat removal was ineffective and that core conditions are continuir.g to degrade. Per ERPIP-802,13OO'F corresponds to clad damage on the order of 274. In orter to provide a discriminator from the FCB2

loss" condition (1EOO'F),
s temperature of 13OO'F and increasing is used hem. Calvert Ctrffs EAL Basis Document B31 June 15,1993
RSSION PRDOLLT BARRER DEGRACMTON Thus, the Potential Luss EAL is written as:
~~, ~s l Valid Core Ext Thermocouple Readings GREATER THAN 1300 'F AND increasing~~
~~/ \ Valid means that the thermocouple channel [s] are considered to be operable in accordance with the Technical Specihcations. Soume Documents / References / Calculations:~~
~~1. Emergtocy Response Ran implementation Procedures~~
~~ERMPOD2, Core Damage Assessment Using Core Exit Thermocouples~~
2. CEN152. Emergency Pmcedure Guidelines

3. Emergency Operedng Procedures
~~EDPa. Functional Recovery Pmcedure~~
~~4. Abnormal Operating Procedures ,~~
~~ADP38, Abnormal Shutdown Cooling Condtions Attachment 14. RCS Levels~~
~~5. Updated Anal Safety Ana>/ sis Report~~
~~UFSAR Section 7.5.9. Inadequate Core Cooling instmmentation~~
~~( 1~~
~~'l )~~
l 1 l l l I Cahert Cirffs EAL Basis Document June 15,1993 i B.32
RSSION PRODUCT BARRER OEGRA[M'nON Calvert Chffs Ememence Action Level: CNB3 Radiadori NUMARC Ememency Action Level: Containment 5 Signifcant Radioacme inventory in Containment
Loss- Not Applicable

Potantial Loss - Containment Rad Monitor Reading GREATER THA,I (SiteSpecife) R/hr s
Containment 7 Other (Sce Specifc) Indications NUMARC Generic Basis: [ Containment 5) The (Stespecific) reading is a value which indcates significant fuel damage well in excess of the EALs associated with < loss of both Fuel Clad and RCS Bamers>. As stated in <NUMARC/NESP-OO7> a major release of radioactrvey mquiring offsite protectue actions from com damage is not possible unless a major failure of fuel cladding allows radioactive material to be released fmm the core into the reactor coolant. Regartfless of whether containment is challenged, this amount of actury in containment, if released, could have such severe consequences that it is prudent to treat this as a potential loss of containment, such that a General Emergency declaration is warranted. NUREG-1228. Source Estimations During incident Response to Severe Nuclear Power Plant Accidents, indcates that such conditions do not exist when the amount of clad damage is less than 2CTE. Unless them is a (site-specife) analysis justifying a higher value, it is recommended that a radiation monitor mading corresponding to 20% fuel clad damage be specifed here. <Thus, this EAL cormsponds to loss of both the fuel clad and RCS baniers with Potential Loss of the Containment barrier, and would msult in declaration of a General Emergency >. Them is no
~~Loss" EAL associated with this item.~~
(Containment 7) This < Generic > EAL should cover other (sitespecifc] indications that may unambiguously indicate loss or potential loss of the containment bamer, including indications from area or ventilation monitors in the containment annulus or contiguous buildings Plant-Specific information: Entry into EOP 5 (Loss of Coolant Accdent) or EOP-8 (Functional Recovery Pmcedum) would be made following a LOCA. As part of the mquired actions in these procedures, a check is made of radiation levels extemal to the containment is made to assure that containment bypass has not occurmd. The location of such a leak is indcated by sump alarms, mom level alarms and area RMS alarms. If a signifcant leak bypassing containment existed that could not be isolated, then the acceptance criteria for radiation levels extemal to containment could not be met and this would indicate a CNTMT toss. Thus, the Loss EAL is written as: EOP-5, Loss of Coolant Amiant, Or EOFa. Functonal Rectuary Pmcadure, is implemented AND Radiation Levels Extemal to CNTMT Can NOT Meet Acx:eptance Criteria Can NOT is used because the final safety function status is of concem, not merely the inability to meet certain intermesate status check condtions. Potential Loss EALs 1 and 2 cddress signifcant radioactbe inventory in containment. The plantspecife containment radiation values wem determined from ERPIP-801, assuming 20% fuel clad damage. This pmcedure Calvert C!rffs EAL Basis Document B33 June 15,1993
RSS10N PF00UCT BARRER MGRADATON can be used to determine the containment radiauon monitor readings resuting from 20% fuel clad failure using -I Attachment 2 and assuming no power cx>rvectaon. The radiation rnonitor reading (1-Rb5317A & B. 2-RL5317 A & B) corvesponding to 20% fuel clad failure at 2 hours after shutdown is about 14,000 rem /h (140 Gray /h).
~~. Thus, Potential loss EAL 1 is wntren as: ,~~
l Valid RL5317A/B Reading of AT LEAST 14,000 mm/h Within 2 Hours After Reactor Shutdownl Valid means that the applicable radiation monitoring channel (s) are considered to be operable in accordance with , the Technical Specifications. The EAL uses the value of 2 hours after the initiating event (assumed to closey correspond to the time of reactor shutdown) for simplicity in pmsentation to the Shift Supervisor acting as the Site Emergency Coordinator (MC). l The two hour point was also picked because it allows ample time for transfer of the SEC duties to outside the Control Room. 9 Technical support personnel can also use ERPIP.801. -802. -803, and -804 to determine core damage. Thus, Potential Loss EAL 2 is wntten as: AT LEAST 2CFK Clad Damage As Determined From Core Dama0e Assessment Potential Loss EAL 3 addresses conditions where leakage outsida containment is detected. As part of EDP-5, ! Loss of Coolant Accident, the operator is instructed to review potential leak paths from the RCS to outside the Containment and isolate such paths, if possible. Actions to be periormed include verifying:
"
Letdown line isolation are shut

RCS sample isolation valves are shut

Leakage into the Component Cooling System is not occurring Existence of leakage outside containment would indicate that the plant systems were not performing in ,
accordance with the design basis for containment isolataon. Therefore, it is appmpriate to classify this condltion as a Potential Loss. ; Thus, Potential Loss EAL 3 is wrxten as: EDP 5, Loss of Coolant Arwiant, is implemented AND LOCA is l'OT Occurring Within the CNTMT As indicated by Aux Building Sump Alarms or Aux Building RMS Alarms if the leakage were significant and could not be isolated, then the acceptance criteria for radiation levels levels extemal to containment could not be met and the 1.oss condition specified above would exist. Soume Documents / References / Calculations: , i
~~1. Emergency Response Plan implementation Pmcedures ;~~
ERPIP-801. Core Damage Assessment Using Containment Radiation Dose Rates

ERPIP-802, Core Damage Assessment Using Core Exit Thermocouples

ERPIP-803, Core Damage Assessment Using Hydmgen -)

ERPIPBO4, Core Damage Assessment Using Radiological Analysis of Samples <
~~2. Emergency Operating Pmcedures~~
EDP-5, Loss of Coolant Accident i

EDP-8, Functional Recovery Procedure i
Calvert Cliffs EAL Basis Document B:34 June 15,1993 I I
RSSON PRODUCT BARRER [EGRADATDN Cakert Cliffs Emergney Action Levet CNB4 Coolant Leakage NUMARCJmemency Action Levet Containment 3 Containment isolation Vake Status After Containment isolation
Loss - Vake(s) NOT Closed and Downstream Pathway to the Envimnment Exists ,

Potentialloss - Not Applcable Containment 4 SG Secondary Side Release Weh PnmarytchSecondary Leakage ,

Loss - Release of Secondary Side to Attrosphem with Primary to Secondary Laakage GREA'IER THAN TS Allowable

Potential Loss - Not Applcable NUMARC Generic Basis:
(Containment 3] This < Generic > EAL is intended to addmss incomplete containment isolation that al lows dimct miease to the envimnment. It mpresents a loss of the containment banier. Them is no Potential Loss EAL associated wth this tem. (Containment 4] This < Generic > EAL addmsses SG tube ruptures. Secondary side releases to atmosphere include those fmm the condenser air ejector, atmospheric < steam > dump vakes, and main steam safety vakes. For smaller bmaks not ' exceeding the normal charging capacity threshold in Potential Loss <EAL RCB4, Coolant Leakage >, this EAL resuts in an Unusual Event < declaration under C BU2, RCS Laakage>. For larger breaks, <EAL RCB4 Potential Loss > would resut in an Alert. For < larger spectrurre SG tube ruptures < >, this < Loss > EAL would exist in conjunction with <EAL RCB4, Coolant Laakage or Loss EAL RCB2, Temperature;> and would resut in a Srte
~~<E>mergency. Escalation to General Emergency would <then> be based on~~
~~Potential Loss" of the Fuel Clad Barrier.~~
Plant-Soecific Information: Loss EAL 1 addresses contianment isolation vake status. t is written in language that faciltates operating staff mcognition. In accortlance with TS 3.6.1.7 and TS 3.6.4, the containment purge isolation valves are not in operation for Modes 1,2,3, & 4; they are closed. The ony time the vakes are operation is Mode 6 under administrative contml Thus, Loss EAL 1 is wrtten as: Leakage Pathway Exists From inside CNTMT to Outside CNTMT Existance of a leakage pathway may be determined by radiation monitoring, physical observation, or by control room valve indcations. Larger spectrum steam generator tube ruptures of concem for Loss EAL 2 will resut in entry into EDP6, Steam Generator Tube Rupture. EDP6 requires that ruptured SG water level be maintained between +30 inches and-170 inches, if the water level cannot be maintained within these limbs, entry into EDP-8, Functional Recovery Procedure, is then made. EOP-8 has a wider allowable acceptance band for ruptured SG water level, between
~~+50 inches and -170 inches.~~
O Calvert Cirffs EAL Basis Occument 8.35 June 15,1993
RSSION PRODUCT BARRER TGRADATDN If the ruptumd steam generator water level could not be maintained below +50 inches, this would indicata a larger spectrum steam generator tube rupture wth potential to overfill the ruptumd SG. AdditionalS in order to have a release to atmosphere following isolation of the affected SG in accordance with EDP4, the affected SG fm N steam pmssum would need to exceed the minimum steam generator safety velve set pressum or the atmospheric dump valve minimum pressum. At Calvert Citfs, this cormsponds to about 900 psig. Thus, Loss EAL 2 is wntten as: SG Tube Ruptum in Pmgmss AND Both of the Following:
Affected SG Level Can NOT Be Maintained LESS THAN +50 inches AND e Affected SG Pressure GREATER THAN 900 PSIG l Can NOT is used to indcate a larger spectrum SG tube ruptum that can result in overfilling the effected SG.
~~Source Documentg/Refemnces/Calculatiqris:~~
~~1. Technical Specifcauons~~
TS 3.4.62, Reactor Coolant System Laakage

TS 3.6.1.7, Containment Purge System

TS 3.6.4, Containment isolatbn Vakes
~~2. Abnormal Operating Procedums~~
~~AOP4A. Excesske Reactor Coolant Leakage i~~
~~3. Emergency Operating Procedures~~
EDPO Post-Trip immediate Action

EDP-6. Steam Generator Tube Ruptum o EDPB, Functional Recovery Procedum
~~4. Updated Final Safety Anatysis Report~~
Chapter 6 Engineered Safety Featums

Section 9.1, Chemical and Volume Control System O
Cabert Ctrffs EAL Basis Document June 15,1993 B.36
i e FISSIDN PRODUCT BARRIER DEGRADATIDN Calvert CI:ffs Emeroence Action Level: r CNB5 Pmssure ( NUMARC Ememency Action Level: Containment 2 Containment Pressure
Loss - Rapid Unexplained Decmase Following initial increase DR Containment or Sump Level NDT Consistent with LOCA Condtions

Potentialloss -(Site-specifc) PSIG and Indreasing DR Explosive Mixture Exists DR Containment Pmssure GREATER THAN Cuntainment Depressurization System Setpoint Wch LESS THAN Dne Full Train of Depressurization Equipmetn Operating Barrier. Containment NUMARC Generic Basis:
Rapid unexplained loss of pressum (i.e, not attributable to containment spray or condensation effects) following an initial pmssure increase indicates a loss of containment integrity. Containment pmssum and sump levels should increase as a result of the mass and energy miease into containment fmm a LOCA. Thus, sump level or pressure not increasing indicates containment bypass (V-sequence) and a loss of containment integrity. The (site-specific) PSIG for potentialloss of containment is based on the containment design pmssure. Existence of an explosive mixtum means a hydmgen and oxygen concentration of at least the lower deflagration limit curve exists. The indcations of potential loss under this EAL corresponds to some of those leading to the RED path in EAL CNB1 above and may be declared by those sites using CSFSTs. <T>his < Generic > EAL is primarily a discriminator between Site <E>mergency and General Emergency representing a potentialloss of tne third barrier. (] The second potential loss EAL represents a potential loss of containment in that the containment heat (") removal /depressurization system (eg., containment sprays, ice condenser fans, etc., but not including containment venting strategies) is either lost or performing in a degraded manner, as indicated by containment pressum greater than the setpoint at which the equipment was supposed to have actuated. Plant-Specific in'ormation: The Calvert Cirffs Loss EALs correspond directly to the NUMARC EAL Because it is diffc' ult to determine whether pressure and sump level response is consistent with expected, and the other Containment Loss EALs address containment response in a way that is observable by the operations staff, the second condition specified in the generic EAL is not used at Calvert Cirffs. Thus, Loss EAL 1 is written as: Rapid Unexplained CNTMT Pressum Decrease Following initial increase The design pressure for the Cakert Cliffs containment is SD psig. Pmper actuation and operation of the containment spray system when required maintains containment pressure below its design pressum following LOCA or secondary side break inside containment. Thus, Potential Loss EAL 1 is wntten as: CNTMT Pressure of AT LEAST SD PSIG and increasing The EAL uses the lower limit of flammabilty of hydmgen in air, i.e,4% hydmgen concentration. However,4% l corresponds to the explosive mixture condtion spec /ied by NUMARC. t Calvert Cirffs EAL Basis Document 8:37 June 15,1993 ;
FISSION PRODUCT BARRER DEGRADATON Thus, Potential Loss EAL 2 is written as: CNTMT H, Concentration of AT LEAST 4.0% Source Documents / References / Calculations:
~~1. Technical Specifcations~~
~~TS 3/4.6, Containment Systems~~
~~2. Emergency Operating Procedures~~
~~EOP-8. Functional Recovery Procedum~~
~~3. Updated Final Safety Analysis Report~~
~~Chapter 6 Engineemd Safety Features I\~~
.L CaNert Cliffs EAL Basis Document B:38 June 15,1993
i RSSION PRODUCT BARRER DEGRADATON ! _ . . . _ _ j Calvert Chffs Ememency Action Level: V q CNB6 SEC Judgement NUMARC Emeroency Action Level: Containment 8 Emergency Dimator Judgement i ((UMARC Genene Bass: l This EAL addmsses any other factors that am to be used by the <iEC)in determining whether the Containment bamer is lost or potentialy lost. In addition, the inability to monitor the barrier should also be incorporated in this EAL as a factor in <SEC> judgemere that the berrier may be considered lost or potentialty lost. (See also
~~<1C EG1, Prolonged Station Blackoutp for additional information.) l Plant-Specific information:~~
Per the Emergency Response Plan, the Site Emergency Coortlinator (SEC) is the title for the emergency director l J function at Calvert Chffs. SEC considerations for determining whether any barrier Loss er Potential Loss include imminent degradation, bamer monitoring capabiltty, and dominant accdent sequences. This information is included on the Rssion Pmduct Bamer reference page which is to be reviewed by the SEC before using the Fission Pmduct Barrier Table. Anticipated degradation of ANY Barrier within 2 hours based on a projection of current safety system performance is considered to be imminent Bamer degradation. This must be considered by the SEC for timely declaration of a General Emergency. The term imminent mfests to the inability to reach final safety acceptance before completing all checks. \ De creased bamer moniconna abilty from loss of/ lack of reliable indicators must also be considered by the SEC when judging whether a Bamer may be Lost or Potentially Lost. This assessment should also include instrumentation operabilty concems, readings fmm portable instrumentation, and consideration off-site monitoring msults. Dorrinant accident sequences will lead to degradaten of all Bamers. The SEC should also consult Station Blackout and ATWS ICs, as appmpriate, to assure timely emergency classifcation declaration. Thus, the EAL is wntren as: Condtions Which in the SEC's Judgement Indicate Loss or Potential Loss of the CNTMT Barrier Based on:
Imminent Bamer Degradaten Due to Safety System Perfortnance

Degraded Abilty to Monitor Bamer Source Documents /Refemnces/ Calculations:
~~1. Emergency Response Plan L~~
4 Catvert Chtts EAL Basis Document June 15,1993 B:39
O EQUIPMENT FAILURE O O
EQUIPMENT FAILURE E_m_e. rgency Classifcation Level: UNUSUAL EVENT C A_pplicable Doerational Modes: 4. 5, 6 Catvert DMs initiatino Condition: GU1 Unplanned Loss of Any Funcdon Needed to Maintan Cold Shutdown NUMARC Recocnition Cateoory System Malfunction NUMARC Initiatino Condition: Not Applicable Barrier Not Appleable NUMARC Generic Guidance: None Bant-Specific information: Unplannedis used to preclude the declaration of an emergency where a component or system has been mmoved intentionalty fmm service (e.g., for maintenance and testing). In order to maintain the anticipatory overall philosophy of the NUMARC EAL methodology and to assure that pmcursom to shutdown accidents are appropnately classifed, Calvert Giffs has added this initiating Condition in the Unusual Event classifcation. NUREG 1449 raises concems regarding inadvertent Cnticality Events during shutdown, in its regulatory analysis of the NUMARC methodology, NRC noted that there is a likelihood that the results of ongoing risk studies relating to shutdown (e.g., NUREG-1449) may necessitate revision of both existing NRC EAL guidance and the new NUMARC guidance as well. Thus, Calvert Orffs has added this C that precursor events of concem are appropriately addressed and to better assure that the NUMARCbased methodology was complete befom its implementation at Calvert Drffs. Per the Technical Specifcations, the functions required to be operable during Cold Shutdown and Refueling modes and are associated with maintaining required shutdown condtions (temperature, pmssure, and subcriticality) are:
Reactivity Control System (TS 3.1)

Coolant Loops and Coolant Circulation (TS 3.4.1,3.9.8)

Safety Valves (TS 3.42)

Deerpressure Pmtection Systera (TS 3.4.9.3)

DrwSite Power Sourtes (TS 3.8)
AC and DC power systems availability are separately addmssed under the Loss of Power Event Category. Thus, J these are not addressed under this initiating Condition. RCS leakage (e.g., requiring use of the Charging /HPSI Subsystems or msutting from Overpressure Protection Syctem malfunctions) is addressed by C BU2, RCS Leakage, and the Radicad:ivity Release Cs related to uncovery of irradiated fuel. Boration systems are addressed by maintaining requimd Shutdown Margin (SDM) as discussed below. Loss of SDC [which is requimd by Technical Specifcations) includes loss of shutdown cooling support functions such as Component Cooling Water that are required to remove heat from the Shutdown Cooling heat exchangers. Under the conditions of concem, ADPGB, AbnorYnal Shutdown Cooling Conditions, would be entered. I June 15,1993 Calvert Orffs EAL Basis Document 0:1 )
EQULPMENT FAILURE Ememency Classification Level. UNUSUAL EVENT apphcable DoerationalMod s: 1.2,3,4 \_./ Raivert CI'fs indatinc Condi;ioD: GU2 Unplanned Loss of Most or All Safety Systam Annurnators for GREATER THAN 15 Minutes NUMARC Recoggdon Catego_ry System Malfunction NUMARC Initiatino Condition: SU3 Unplanned Loss of All Safety System Annunciators for Gmatar'Ihan 15 Minutes Barner: Not Applicable NUMARQ Generic Basis: This IC and its associated < Generic > EAL em intended to recognize the drfficulty associated with monitoring changing plant condtions without the use of a major portion of the annunciation or indication equipment. i Recognbon of the availabilty of computer based indcation equipment is considered (SPDS, plant computer, etc.). Unplanned loss of annunciators or indicators excludes scheduled maintenance and testing activities. Compensatory nonetarming indications in this context includes computer based informaton such as SPDS. This should include all computer systems available for this use depending on specific plant design and subsequent retmfts. Quantifcation of "most* is arbitrary, however, it is estimated that if appmximately 75% of the safety systems annunciators or indcators are lost, there is an incmased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use the value as a judgement threshold for determining the severity of the plant conddons. This judgement is supported by the specife opinion of the Shift Supervisor that addtional operating personnel will be required to pmvide increased monitoring of system operation to safety operate the unit (s). t is further recognized that most plant designs pru/cle redundant safety system indcotion fmm separate uninterruptible power supplies. While failure of a large portion of annunciators is more liksty than a failure of a large portion of indcations, the concern is included in this EAL due to the drffcuty associated with the assessment of plant condtions. The loss of specife, or several, safety system indicators should remain as a function of that specific system or component operabiley status This will be addressed by the specific Technical Specifcation. The initiation of a Technical Specification imposed plant shutdown related to the instrument loss will be reported via 1D CFR SD.72. If the shutdown is not in compliance with the Technical Specifcation action, the Unusual Event is based on <!C QU4, Inabilty to Reach Required MODE Within Technical Specification Umits>. (Srte-specific) annunciators or indcators for this EAL include those ioentified in the Abnormal Operating Procedures, in the Emergency Operating Pmeedures, and in other EALs (e.g., area, pmcess, and/or effluent rad monitors, etc.) Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Due to the limited number of safety systems in operation during cold shutdown, refueling, and defueled modes, no IC is indcated during these modes of operation. Catvert Citffs EAL Basis Document G.3 June 15,1993
EQUIPMEr# FAILURE Thus, EAL 1 is wrtten as: Entry into ADP-38, Abnormal Shutdown Cooling Conditions, is Requimd For GREATER THAN 15 Minutes ( Required means that entry into the Abnormal Operating Pmcedum is neither optional nor memh suggested, but rather imperative based on existing condtiors. The Cold Shutdown and Refunling Modes are defined by specific plant conditions - mactMty condition (K,,) and coolant temperature. Maintenance of the ebilty to mmove com decay heat addmsses the coolant temperatum criteria. The EALs addmssing mquired subcrtcalty conditions for operation in modes 5 and 6 am missing fmm the NUMARC EALs. Per Technical Specifcation Table 1.1 Operational Modes, the requimd SDM is K,,, less than O.99 for Mode 5 and K,,, not mom than U.95 for Made 6. Per Technical Specircation 3.9.1, the minimum bomn concentration requimd during refueling mode is at least 2300 ppm. Under the condeions of concem ADP-1 A would be entered. Thus, EAL 2 is wntten as:
~~-e-,~~
~~Entry into ADP-1 A, inadsartent Bomn Dilution, is Required AND Shutdown Maqjin NOT Maintained Source Documents / References / Calculations:~~
1. Technical Specifications

2. Abnormal Operating Pmcedures
AOP-1 A, inadvertent Bomn Dilution

AOP-3B, Abnormal Shutdown Cooling Conditions
3. NUREG 1449, Shutdown and Low-Power Operation at Commerr.ial Nuclear Power Plants in the United States, Draft for Comment February 1992 4 Regulatory Anahsis - Revision of Regulatory Guide 1.1D1 to Accept the Guidance in NUMARC/NESPOO7, Rev. 2 as an Attemative Methodology for the Development of Emeqpncy Amion Levels Catvert Ctrffs EAL Basis Document Q2 June 15,1993
EQUIPMENT FAILURE The Unusual Event will be escalated to en Alert ;f a transi nt is in progress during the loss of ennunciation or indication. Rant-Soecrfic information: The EAL is based on NUMARC. INPO SER 16-92 was rewewed and determined not to apply to Calvert Cliffs Annunciator design. Thus, the EAL is wrnten as: Unplanned Loss of 75% of Main Contd Boart! Annunciators for GREATER THAN 15 Minutes Soume Documents / References / Calculations:
~~1. Abnormal Operating Pmcedures~~
~~ADP-7J. Loss of 120 Volt Vital AC or 125 Volt Vital DC Power~~
2. Updated Rnal Safety Analysis Report

3. INPO Significant Event Report (SER) 16-92. Loss of All Annunciation When Computer Lost With Annunciators CaWrt Cirffs EAL Basis Document 0:4 June 15.19P.3
~~EQUIPMENT FAILUE Ememency Classifcation Levet: UNUSUAL EVENT~~
~~@plicabigSperational Modes: ALL Calvert Cirffs initiatino Condcion:~~
003 Unplanned Loss of All OnGita or OffGita Commurucations Capabirrties NUMARC Reco_g. n ition Ca_tegory System Malfunedon NUMARC Initiatino Condition: SU6 Unplanned Loss of All OrsSte or Off6ite Communications Capabilities Barrier: Not Applicable NUMARC Generic Basis: The purpose of this BC and its associated < Generic > EALs is to mcognize a loss of communications capability that either defeats the plant operations staff abilty to perform routine tasks or the abilty to communicate problems with offsite authorcies. The loss of offsite communications capabilcy is expected to be signifcarrJy more comprehensive than that addmssed by 10 CFR 50.72. (Scespecife list) onsite communications loss must encompass the loss of all means of routine communications (i.e., phones, sound powered phone systems, page party system and rsdios/walkie talkies). [Sitespecific list) offsite communications loss must encompass the loss of all means of communications with off-
~~- site authorities. This should include the ENS, Bell Unes, FAX transmissions, and dedcated EPP phone systems.~~
This EAL is intended to be used only when exttsordinery means are being utilized to make communications ' possible (mlaying of information from radio transmissions, individuals being sent to offsite locatons, etc.). Plant-Specific information: A communication system with multiple redundancy has been prtuded to ensum availability and ease of operation. The communication system consists of six electronic subsystems:
Plant Public Address

Sound powered phones for plant use

Commercial Telephone

Sound powered phones for emergency use

Memwave System

Radio telephone system Thus, EAL 1 is written as:
~~l Loss of ALL On-Site Electmnic Communications Methods Communications with off-site authorities are prtmded by three sted:ronic methods They are:~~
Dedcated Off6ite Agency Telephone

Commercial Telephone

Radio Telephone System EAL 2 is wrtten as:
Loss of ALL Telephone Communications Wch Govemment Agencies A Cahert Cliffs EAL Basis Document CL5 June 15,19S3 : m-
EQUIPMENT FAILURE Source Documents / References /Calculadons:
~~1. Emergency Response Plan~~
~~Chapter 5. Facildes and Equipment; Section !!. Communications~~
~~/ )T L 2. Emergency Response Plan implementation Pmcedures~~
ERPIP 508. Plant Parameters Communicator, EOF

ERP!P 901, Communications Equipment n
v CaNert Cirffs EAL Basis Document Q.6 June 15,1993
EQUIPMENT FAILUFE i Ememency Classification Levet UNUSUAL EVENT /~' Applicable Doerational Modes: 1,2,3,4 . Cakert Chffs initiating Condtion: G04 inabihty to Reach Required MODE Wahin Technical Speci&dde Limas NUMARC Becognition Catecone. System Malfunction , NUMARC Initiatino Condciorl: SU2 Inabilcy to Reach Requimd Shutdown Within Technical Specifcetion Umits Darrier Not Appleable NUMARC Generic Basis:
~
Umcing Condtions of Operation (LCOs) require the plant to be brought to a required shutdown mode when the ! Technical Specifcation required configuration cannot be restored. Depending on the cimumstances, this may ; or may not be an emergency or precursor to a more severe condition, in any case, the initiation of plant ! shutdown required by the site Technical Specifcations requires e one hour report under 10 CFR 50.72 (b) Non-emergency events. The plant is wthin its safety envelope when being shut down within the allowable action L statament time in the Technical Specifications. An irnmediate Notdicaton of an Unusual Event is required when the plant is not brought to the required <operationa> mode within the allowable action statement time in the j Technical Specifcations. Declaretion of an Unusual Event is based on the time at which the LEOspeedied acnon statement time period elapses under the alta Technical Specifememns and is not related to how long a condition may have antad Other required Technical Specifcation shutdowns that invoke precursors to more serious ' events are addressed by <lectreal. Equipment Failure Fission Pmduct Barrier Degradation, and Other Hazards > j ICs. t 3 Plant-Specific Information: LCOs, their associated action statements, and applicable time frames for placing the unit in a shutdown mode l are found in the Cabert Cliffs Technical Specifcations, Section 3D.3. When an LCO is NOT met, except as ' pmvided in the associated action requirements, then other action requirements apply as stated in Applicabilty, ; Section 3/4D. ' l Thus, the EAL is wntten as: ' Unit Can NGT Be Bmught to Required Mode Wthin Applicable LCO Action Statament Time Umits Source Documents / References / Calculations: 1, Technical Specifcations )
~~TB 3/4D Applicabilty- Umiting Condtion for Operation 1~~
)
O ' Cabert Cliffs EAL Basis Document gy June 15,1993 i l m -
EQUIPMENT FAILURE Em_ergency Classification Level: ALERT Applicable Operational Modes: 1, 2 Catvert Cliffs hdatino Condition: GA1 Failum of Autornado Reactor Trip l NUMARC Recognition Oseco_ry System Malfunction NUMARC Indatino Conddon: SA2 Failuru of Reactor Protection System Instrumentation to Complete or inuate an Automatic Reactor Scram Once a Reactor Protection System Setpoint Has Been Exceeded and Manual Scram Was Successful Bagier Not Applicable NU. MARC Generic Basis: This conddon indicates failure of the automatic pmtection system to scram the reactor. This condition is more than a potential degradation of a safety system in that a front line automatic protection system did not funcion in response to a plant transient and thus the plant safety has been compmmised, and design limits of the fuel may have been exceeded. An Alert is indicated because conddons exist that lead to potential loss of fuel clad or RCS. Reactor protection system setpoint being exceeded (rather than limung safety system setpoint being exceeded) is specifed here because failure of the automatic pmtection system is the issue. A manual scram is any set of saions by the reactor operator (s) at the reactor control console which causes control mds to be rapid}y insertad into the core and brings the reactor subcrocal(e.g, reador trip button). Failum of manual scram would escalate the event to a Site @mergency. \ Plant-Specific information: Exceeding any of the following setpoints should result in an automatic reactor trip: REACTOR TRIP COINCIDENCE SETPOlt# High Power Level 2/4 Variable High Rate-of-Change of Power 2/4 below 15% RTP 2.6 decade / min. Low Reactor Coolant Row 2/4 above 10% RTP Variable Low Steam Generator Pressure 2/4 670 psig Low Steam Generator Water Level 2/4 10' below top of feed ring High Pressurizer Pressure 2/4 2385 psig Thermai Margin / Low Pressure 2/4 above 10% RTP Variable Loss of Load 2/4 above 15% RTP N/A High Containment Pmssurs 2/4 4 psig Axial Rux Offset 2/4 Variable Thermal Margin /SG Press. Drff. Hi 2/4 above 10% RTP 135 psid s Catvert Chffs EAL Basis Document June 15,1993 G.8
EQUIPMENT FAA.URE Per EON, Post Trip immediate Actions, the operator b to ensure that th reactor has tripped by depressing on3 set of Manual Reactor Trip buttons immediately following any symptoms of a reacor trip. These symptoms include: ( (/ e Reactor Trip alarm Contml Element Assembly (CEA) Circuit Breaker (s) Trip alarms , Rapid Lowering in Reactor Power e
Pmtection Channel Trip elarm

Reactor Pmtecdve System (RPS) Trip Bistable Ughts lic -
Following depression of the reactor trip buttons, the operscor is to venfy that moctor power is decreasing. If these responses can no; be verified, then as part of contingency actions, the cparator is instructed to open the motor generator (MG) set feeder breakers providing power to the Control Eiernent Dnva Mechanism (CEDM). Entry into the Alert emergency classification oxurz sftenever it is determined by the S%ft Superusar that a required automatic reactor trip did not occur, based on the entry conditions into EDPD listed above. It is recognized that EDPO instructs the operator to depress Amnual trip buttons, whether or not a required j automatic reac:or trip actually occurved. However, the failure of a redundant front 4ine automatic protection : system function (i.e., the failure of the RPS to complete a mactor trip following receipt of a trip signa 0 rnee:s the l Alert classification threshold of potentia! substantial degradataon in the level of safety of the plant. This is *. rue i whether or not fuel damage is determined to have occurred. Thus, the EAL is witten as: Automatic Reactor Trip Signal Generated AND Manual Trip Was Required to Trip the Reactor (EOPO, PosteTrip immediate Acions. Reactrvey Contrui, Successfu0 l 1 Source Documents / References / Calculations: \'
~~1. Technical Specifications~~
~~TS 3/4.3.1, Reactor Pmtecdve Instrumen*ation~~
~~2. Emeq;ency Operating Pmcedures~~
~~EDPO, PosteTrip immediate Actions~~
~~3. Updated Final Safety Anatysis Report~~
~~Chapter 7, instrumentation and Control Calvert Citffs EAL Basis Document o.g June 15,1993 l~~
EQUIPMENT FAILURE Emergency Classification Level: ALERT Applicable Operational Modes: 5, S Cakert CMfs initiating Conddon: QA2 Inability to Maintain P(ant in Cold Shutdown NUMARC Recognition Category System Malfunction NUMARC Initiatino Conddon: SA3 Inabilty to Maintain Plant in Cold Shttdown Barrier' Not App! cable r NUMARC Generic Basis: This C and its associated EAL addmss complete loss of functions required for com cooling during mfueling and cold shutdown modes. Escalation to Site <E>mergency or General Emergency would be via <RadioactAty Release or SEC Judgement > Cs. For FWRs, this C and its associated EAL are based on concems raised by Generic Letter 88-17 Loss of Decay Heat Removal. A number of phenomena such as pmssurization,vortexing, steam generator U4:ube draining. RCS level diffemnces when operating at a mid-loop conddon, decay heat mmoval system design, and level instrumentation pmblems can lead to condtions where decay heat removal is lost and core uncovery can occur. NRC anahses show sequences that can cause core uncovery in 15 to 20 minutes and sevem com damage within an hour after decay heat removal is lost. Under these conddons, RCS integrity is lost and fuel clad g
*) integrity is lost or potentially lost, which is consistent wth a Ste <E>mergency. (Sitespecific) indicators for these EALs are those methods used by the plant in response to Generic Letter 88-17 which include core exit temperatum monitoring and RCS water level monitoring. In addcion, radiation mon'tor readings may also be appropnate as indicators of this conddon.
Uncontmiled means that system temperature increase is not the result of planned actions by the plant staff. The EAL guidance related to uncontmiled temperature rise is necessary to preserve the anticipatory philosophy of NUREG-0654 for events starting tmm temperatures much lower than the cold shutdown temperature limit. Escalation to the Site <E>mergency is by 41adioactivity Releass> Cs. Mubonit stations with shared safety functions should further consider how this C may affect more than one unit and how this may be a factor in escalating the emergency class. P! ant-Specific Information: Per Calvert Cliffs Technical Specifications, the functions required to be operable during Cold Shutdown and Refueling modes and am associated with maintaining required shutdown conddons (temperature, pressure, and subcrescalty) are:
Reactivity Contml Systems (TS 3.1)

Coolant Loops and Coolant Cimulation (TS 3.4.1,3.9.8)

RCS Safety Valves (TS 3.42)

Dwerpressure Pmtection System (TS 3.4.9.3)
~~+ DrsSte Power Sources (TS 3.8)~~
O l l l Calvert Cliffs EAL Basis Document o.10 June 15,1993
EQUIPMENT FAILURE AC cnd DC power systems av ilability are separately addressed undar the Electncal Event Category. Thus, these are not addressed under this initiating Condtion. RCS leakage (e.g, requiring use of the Charging /HPSI Subsystems or resulting fmm Overpressure Pmtection System malfunctions) am addressed by C BU2, RCS [_ Leakage, and the Radioacivity Release Cs related to uncovery of irradiated fuel Boration systems are addmssed ( by EAL 2 discussed below. EAL 1 is wntzen as: Uncontrolled RCS Temperatum increase of AT LEAST 10'F That Results in RCS Temperature GREATER THAN 200'F This cormsponds to the inabilty to maintain requimd temperature conditions for Cold Shadown. The 10'F threshold was picked to assure that minor cooling interrupthns occumng at the transition between Mode 4 and Mode 5 (that am stready addmssed by GU1) do not result in unnecessary declaration of an Alert. Uncontrolled means that the temperature increase is not due to deliberate oparator action. y condition and reactor Cold Shutdown and Refueling modes are defined by specific plant conditions - core reactivit. coolant temperature / pressure. Maintenance of the abilty to remove core decay heat addresses coolant temperature. The reactkity condtion is addressed by maintenance of required shutdown margin. At the Alert emergency classification. this corresponds to assuring that the reactor is not crtcal. Thus, EAL 2 is wntzen as: Inadvertent Crcicalcy as Determined by Valid Wde Range Logarithmic Channel Indcations inadvertent means accidental or unintentional, e.g., the event occurved because procedures were not stncdy adhered to. /~ Valid means that the indication is from instrumentation determined to be operable in accordance with the ( Technical Specircations or has been verifed by other independent methods such as indications displayed on the control panels, reports fmm plant personnel, or radiological survey results. Sourte Documents / References /Celculations:
1. Technical Specifications

2. Abnormal Operating Pmcedures
~~ADP.GB, Abnormal Shutdown Cooling Conditions :~~
1
~~3. NUREG 1449. Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States, l Draft for Comment, February 1992 l~~
i i Cakert Cirffs EAL Basis Document Ct11 June 15,1993
EQUIPMENT FALURE Eme_rgency Classification Level: ALERT r' Appli_ cable Operational Modes: 1,2,3,4 i ( Catvert Cfffs initiatino Condition: QA3 Unplanned Loss of Safety Systern Annurnators With Transient in Progress NUMARC Recoonition Catego_ry System Malfunction NUMARC inciating_Condtion. SA4 Unplanned Loss of Most or A!! Safety System Annunciation or indication in Contml Room Wth Either (1) a Signifcant Transient in Pmgress, or (2) Compensatory Non Alarming indcators em Unavailable BargeC Not Applicable NUMARC Generic Basis: This IC and ts associated < Generic > EAL are intended to recognize the diffculty associated with monitoring changing plant condtions with the use of a major portion of the annunciation or indication equipment during a transient. Recognition of the availabilty of computer based indcation equipment is considemd (SPDS, plant computer, etc.). Ranned loss of annunciators or indicators includes scheduled maintenance and testing activities. Quantification of most is arbitrary, however, it is estimated that if appmximatey 75% of the safety systems :
, annunciators or indicators am lost, there is an increased risk that a degraded plant condition could go undetected. It is not intended that plant personnel perform a detailed count of the instrumentation lost but use
(% the value as a judgement threshold for determining the seventy of the plant conditions. This judgement is supported by the specific opinion of the Shift Supervisor that addtional operating personnel will be required to pmvide increased monitoring of system operation to safey operate the unit (s). It is further recognized that most plant designs provide mdundant safety system indication from separate uninterruptible power supplies. While failure of a large portion of annunciators is mom likey than a failure of a large portion of indcations, the concem is included in this EAL due to the drffculty associated with the assessment of plant condtions. The loss of specife, or several, safety system indcators should remain as a functinn of that specific system or component operability status. This will be addressed by the specife Technical Specifcation. The initiation of a Technical Specifcation imposed plant shutdown related to the instrument loss will be reported via 10 CFR 50.72. If the shutdown is not in comp!!ance with the Technical Specifcat:on oction, I the Unusual Event is based on <.C QU4, inability to Reach Required MODE Wohin Technical Specifcation Limits >. , (Sitaspecific) annuncistors or indicators for this < Generic > EAL must include those identified in the Abnormal Operating Procedures, in the Emergency Operating Pmoedures, and in other EALs (e.g., area, process, and/or effluent rad monitors, etc.) Significant transient includes response to automatic or manualy initiated functions such as screms, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10% or l greater. l l Compensatory nonetarming indications in this context include computer based information such as SPDS. This should include all computer systems available for this use depending on specific plant design and subsequent retrefts. If a major portion of the annunciation system and all computer monitoring <both> are unavailable ,
~~<such> that the addtional operating personnel are requimd to monitor indcations, the Alert is required.~~
{ k ! Catvert Chffs EAL Basis Document Q:12 June 15,1993 i _ - - .___ a
EQUIPMENT FALURE I Due to the limited number of safety systems in operation during cold shutdown, mfueling, and defueled modes, ! no IC is indcated dunng these modes of operation. ! 1 This Alert will be escalated to a Site <E>mergency if the operating cmw cannot monitor the transient in pmgmss. Plant-Specific Information: Compensatory non-alarming indcations include the Safety Parameter Cicplay System (SPDS) and the plant computer. Thus, the EAL is wrtten as: Unplanned Loss of 75% of Main Control Board Annunciators AND EITHER of the Following:
Plant Transient in Progress

SPDS Dr Plant Computer NOT Available Transient means a condition that is:

Be/ond tha expected steadtstate fluctuations in temperature, pressum, power level, or water level, and

Beyond the normal manipulations of the Control Room operating crtw, and

Expected to require actuation of fastecting automatic control or protection systems to bring the reactor to a new safe, steady-state condition.
Escalation to Site Emergency would be based on plant transient response, occurrence of other equipment malfunctions requiring operator actions outside the control room, or loss of additional monitoring instrumentation (such as ICC instrumentation) requimd to deterTnine plant condcions. Source Documents /Refemnces/ Calculations:
~~1. Abnormal Operating Procedures b~~
kl
~~ADP-7J. Loss of Vtal 120V AC or 125V Vtal DC Power~~
~~2. Updated Final Safety Anatysis Report i~~
O Calvert Cliffs EAL Basis Document 0.13 June 15,1993
EQUIPME?# FAILURE 1 l Emeroency Classification Level: SITE EMERGENCif Applicable Operational Modes: 1, 2 Calvert Cir'fs initiatina Condition: OS1 Failure of BOTH Automatic AND Manual Raactzr Trip NUMARC Rec _ognition Categog System Malfuncdon NUMARC initiatino Condition: SS2 Faibre of Reactor Pmtection System Instrumentation to Complete or initiate on Automatic Reactor Scram Once a Reactor Protection System Setpcint Has Been Exceeded and Manual Scram Was NOT Successful Sartiez Not Applicable NUMARC Generic Basis: Automatic and manual scrams are not considered successfulif action away fmm the reactor control console was requimd to scram the reactor. Under these condtions, the reactor is producing more heat than the maximum decay heat load for wtiich the safety systems are designed. A Site <E>mergency is indicated because condtions exist that lead to imminent loss or potential loss of both fuel clad and RCS. Although this C may be viewed as redundant to the Fission Product Barrier Degradation IC, its inclusion is necessary to better assure timely recognition and emergency response. Escalation of this event to a General Emergency would be via Fission Product Bamer Degradation or
~~<SEC Judgement;> Cs.~~
t Plant-Specific e Information: EOPO, PostTrip immediate Actions, are described under C GA1, Failure of Automatic Reactor Trip. As stated under GA1, entry into the Alert emergency classification occurs whenever it is determined by the Shift Supervisor that a required automatic reactor trip did not occur, based on the entry condtions into EOPO Entry into the See Emergency is made consistent with EOPO pmcedural mquirements and so cormsponds to Dg; satisfying the reacthey control creeria of EDPO. This means that both automatic and manual actions were gpp effectke in bringing the reactor suberscal and that entry into EOPE, Functional Recovery Procedure, is required. Thus, the EAL is wntten as: , EOPE, Functional Recovery Pmcedure, is implemented per EDPO, Post Trip Imnsdate Actions, Rasctrvey Control Can NOTis used because the abilty to meet the final acceptance criteria is the appmpriate concem, not whether intermediate acceptance criteria are not being achieved at one point in time. Equrge Documents / References / Calculations:
~~1. Emergency Operating Pmcedures~~
EDPO, PostTrip immediate Actions

EO% Functiona' Recovery Procedure b
( Cakert Cirffs EAL Basis Document g;14 June 15,1993
EQUIPMENT FAILDRE Ememency Classification Level: SITE EMEFENCY Applicable Doerational Modes: 1.2,3,4 Calvert Cliffs initiating n Condtion: GS2 Completa Loss of Function Needed to Acheve or Maintain Hot Shutdown ((UMARC Recoandon Catego_ry System Malfunction NUMARC Initiatina Condtion: SS4 Complete Loss of Function Needed to Achieve or Maintain Hot Shutdown Bamer: Not Applicable e; -s _N_() MARC Generic Basis: This <C and its associated Generic EA!> address complete loss of functions, including ultimate heat sink and t reactivity control, requimd for hot shutdown with the reactor at pressum and temperature. U6, der these conditions, them is an actual major failure of a system intended for protadaon of the public. Thus, declaration of a Site <E>mergency is warranted. Escalation to General Emergency would be via 41adeoactivity Release, Fission Pmduct Bamer Degradation, or SEC Judgement > Cs. MultRanit stations with shared safety functions should further consider how this C may affect more than one unit and how this may be a factor in escalating the emergency class. Plant-Specific information: i
\ Per Calvert Cirffs Technical Specifications, the following functions are required to be operable during Cold Shutdown and Refueling modes and are necessary to maintain Hot Shutdown (Mode 4) conditions (temperature, pmssure, and subcriticalcy):
Reacdvity Control Systems (TS 3.1)

Coolant loops and Coolant Circulation (TS 3.4.1,3.9.8) e ECCS Subsystems (TS 3.52. 3.5.3)

Refueling Water Tank (TS 3.54)

Safety Valves (TS 3.42) e Service Water System (TS 3.7A)

Deerpressure Protection System (TS 3.4.9.3)

Dn Site Power Sources (TS 3.8) e Monitoring instmmentation (TS 333)

Reactor Coolant System Vents (TS 34.13)
AC and DC power systems availability are separately addressed under the Loss of Power Event Category. Thus, these are not addressed under this initiating Condition. The Overpmssure Protection System and Reactor Coolant System Vents are not direcdy related to core cooling and subcreacality functions. Failums of these systems functions resulting are addressed by Fission Pmduct Barrier Degradation Cs. Loss of Monitoring instrumentation is not direcdy mlated to maintaining subcritacalcy and heat removal functions, and therefore is not required to be addressed by this C. 1 ( l i Calvert Cliffs EAL Basis Document Ct15 June 15,1993
EQUIPMENT FAILUE ! l Per ADP-38, Abnormal Shutdown Cooling Conddorn, auxiliary faedwater and atmospheric statm dump capabilty l to at least one SG is necessary to achieve Hot Shutdown conditions under natural circulation condtions. Around the trensition from Mode 3 to Mode 4, the Shutdown Cooling System (SOCS) is typicaly used as the means to remove sensible and decay heat. Once the SDCS is placed in service, the steam generator heat sink capabilty is no longer necessary. Thus, the EAL mflects that neither the steam generators nor Shutdown Cooling are fuly capable of performing heat removal functions. The applicable acceptance criteria for Core and RCS Heat Removal i are shown on the Safety Function Status Checks and am fully explained under the basis information for EAL FCB1, Safety Function Status / Functional Recovery. Thus, EAL 1 is wrtten as: EDP4, Functaonal Recovery Pmcedure, is implemented AND Either of the Following:
Reactivity Contml Acceptance Cnteria Can NOT Be Met

Shutdown Cooling is NOT in Service AND Core and RCS Heat Removal Acceptance Criteria Can NOT Be Met Can NOTis used because the ability to meet the final acceptance enteria is the appmpriate concem, not whether intermediate acceptance cnteria are not being achieved at one point in time.
In service means that the SDCS is in the proper configuration for RCS beat removal (SDCS isolation valves open, LPSI pumps operating, etc.) and is considered " operable" as defined in the Calvert Cidfs Technical Specircations Section 1.6. In order for there to be a path for heat removal between the core and the steam generators or the shutdown cooling system there must be enough RCS liquid inventory to maintain natural circulation. Recent information from the CE Dwners Group indicates that two-phase natural circulation (reflux boiling) works very well and will maintain the RCS between 200 'F and 300 *F. This requires that the RCS water level be below the top of the hot legs. Per ADP-3B, Attachment 14,50' RVLMS Indication corresponds to the middle of the hot leg and is the 5th RVLMS alarm level. Staying above this level (and below the top of the hot legs at the 71
~~level) assures O that, at a minimum, reflux boiling can be maintained.~~
Thus, EAL 2 is wntten as: Zero (0) Indicated Subcooling Margin Determined Using CET Temperatures AND Valid RVLMS Level Indication of LESS THAN 50 inches Per Technical Specification Table 1.1, Operational Modes, the required SOM is K,, less than O.99 for Mode 4 (Hot Shutdown). The existence of a positke startup rate that could not be eliminated by operation of any reactMty control mechanism corresponds to conditions where a major function intended for the protection of the public has failed and themfore meets the threshold for a Sce Emergency classifcation. Sourre Documents / References / Calculations:
1. Technical Specifcations

2. Abnormal Operating Procedures
~~ADP-3B, Abnormal Shutdown Cooling Condtions~~
~~3. Emergency Operating Procedures~~
~~EON 3, Functional Recovery Procedure~~
~~4. Intemal Memorendum, J. R. Hill to R. L Wenderich, Operations Subcommittee Meeting-Trip Report, April 16,1993 O~~
Calvert Cirffs EAL Basis Document G16 June 15,1993
EQUIPMENT FAILURE Ememency Classification Level: SrTE EMERGENCY r3 applicable Operadonal Modes: 5. B t : Qahert Cliffs initiatino Condition: GS3 Loss of Water Lavol That Can Uncover Fuel in the Reactor Vessel NUMARC Initiating Condition: SS5 Loss of Water Level in the Reactor Vessel That Has or Will Uncover Fuel in the Reaczor Vessel Bamer FUEL CLAD NUMARC Generic Basis: Under the cond,tions specified by this C, severe core damage can occur and mactor coolant system pressure boundary intagncy may not be assured.o For PWRs, this C covers sequences such as prolonged boiling following bss of decay heat removal. Thus, declaration of a Site <E>mergency is warranted under the conditions specified by the C. Escalation to a General Errergency is via <Radioactuty Release C RG1, OffEite Dose of AT LEAST 1 REM (EDE+ CEDE) Whole Body or 5 REM (CDE) Thyroid >. Plant-Specific Infonnation: Sequences that can result in uncovery of fuel in the reactor vessel (indirectly by prolonged boiling) include leakage through SG nozzle dams, pipe breaks in the Shutdown Cooling (SOC) System or Chemical & Volume Control [ System (CVCS), or loss of the SDC function. These leakage sourt:es are outside the reactor vessel at most could \ only result in water level decreases to the bottom of the hot leg elevation. This water level decrease would cause loss of SDC suction, trx: ore instrumentation (Cl) penettetaons for Calvert Cliffs are thmugh the vessel head. Thus, these do not have to be considered for this C. A review of attachments to ADP 38, Abnormal Shutdown Cooling Condtions, shows that depending on pmvious j power history and assuming an initial RCS temperature of 140*F, boiling in the core can begin in as little as 7 j minutes following loss of SDC during micWoop operation. ADPGB also shows that under these conditions,without any operator action, core uncovery can begin wthin about BD minutes after loss of SDC. Available methods to restore RCS inventory and to remove core heat include restoring the SDCS, injecting into the RCS from the Refueling Water Tank (RWT) using the HPSI, LPSI, CS or charging pumps, using the steam generators as a heat sink, using the Refueling Pool as a heat sink, aligning a LPSI pump to take suction from the RWT, or wen injecting into the RCS using Safety injection Tanks (SITS). Grven the number of methods to restore inventory, and the amount of time available, it is highly unlikely that this C will be entered. Thus, the EAL is wntren as: ADNB Abnormal Shutdown Cooling Conditions,is implemented AND ANY of the Following Conditions Exist:
~~Altemate Methods for Restoring RCS Inventory Are NOT Effective e Valid RVLMS Reading indicating 0% Level )~~
e Valid CET Reading Indicating Supertiest Conditions O Calvert Cliffs EAL Basis Document g;17 June 15,19S3 _ ____._____._a
EDUIPMENT FA!L1JRE I l l 107 Effective merns that inventory is not being mstored based on availabe operable instrumentation readings ! such as TTs, RVLMS, Hot Leg Level, or fmm decreasing level indications fmm appicable sucion sources such as the RWT, containment sump, or STTs. () ew , Valid means that the indication is fmm instrumentation determined to be operable in accordance with the Technical Specifications or has been venfed by other independent methods such as indications displejed on the control panels, reports fmm plant personnel. or radiological survey results. For example, under condtions whem the GTs and the RVLMS am disconneded to allow mactor vessel head removal, these instrument medings would not be valid. Sou_rpe_Qocuments/ References / Calculations:
~~1. Abnormal Operating Pmcedums~~
~~AOPGB, Abnorinal Shutdown Cooling Conditions O~~
Calvert Cirffs EAL Basis Document CL18 June 15,1993 o
EQUIPMENT FAL1JFE lLm_emency Classification Level: GENERAL FMERENCY epplicab'e Operational Modes: 1 \ Cakert Cliffs intiatino Conddon: OG1 Feilure of BOTH Automatic AND Manual Paarere Trip -AND- ; Exlreme Challenge to the Ability to Cool the Cors { NUMARC Recognition Cateco_ac System Malfunction ; i NUMARC Initiatino Condition: SG2 Failure of the Reactor Pmtection System to Complete an Automatic Scram and Manual Scram was NOT ! Successful and There is Indication of an Extreme Challenge to the Abilty to Cool the Cors ; Barrier: Not Applicable NUMARC Generic Basis: Automatic and manual scrams are not considered successful if action away fmm the reactor contml console is mquired to scram the reactor. l Under the conditions of this IC and its associated < Generic > EAL.tte efforts to bring the reactor subentical have { been unsuccessful and, as a result, the reactor is producing more heat than the maximum decay test load for i which the safety systems were designed. Although there am capabilities away from the reactor control console, such as emergency boration,o the conti ug temperature rise indicates that these capabilities are not effective. l This stuation could be a precursor for a : ore melt sequence. j For FWRs the extreme challenge to the abilty to coul the core is intended to mean that the core exit ; temperatures are at or appmaching 12OO'F or that the reactor vessel water level is below the top of the active j fuel. o Another consideration is the inabilty to initiaHy remove heat during the ear 1y stages of this sequence. For FWRs, ; if emergency feedwater flow is insufficient to remove the amount of heat required by design from at least one steam generator, an extreme challenge should be considemd to exist. o > In the event ether of these challenges exist at a time that the reactor has not been brought below the power associated with the safety system design [ typically 3% to 5% power), a core melt sequence exists. In this ; situation, core degradation can occur rapidly. For this reason, the General Emergency declaration is intended l to be anticipatory of the fission product barrier matrix declaration to permit maximum offsite intervention time. l i ELant40ecific Information: EDP 0, PosteTrip immediate Actions, are described under IC GA1, Failure of Automatic Reactor Trip. As stated under GS1, entry into the Site Emergency classification means that both autorretic and manual reactor trip were ; not effective in bringing the reactor subentcal and that funczional recovery of reactMty contml is required in i accordance with EOFL8 Escalation to the General Emergency is indicated whenever Reactor power is not . decreasing following actions to bring the reactor subcrtcal including automate and manual reactor trip. manually l inserting the control rods, tripping the CEDM motor generator sets or performing emergency boration and there are indications of inadequate core cooling. i i i Calvert Cirffs EAL Basis Document G19 June 15.1993 l
h EQUIPMENT FALURE ! I Thus, the EAL is written ts- t i EDPa, Fundaonal si " mcodure. P is implernented AND Both of the Following:
Reactreity Control Can NOT Meet Acceptance Cnteria AND e Com and RCS Heat Removal Can NOT Meet Acceptance Cntaria f Can NOT is used because the abilty to meet the final acceptance criteria is the appmpriate concem, not whether ,
l intermediata acceptance criteria are not being achieved at any given moment. Source Documents / References / Calculations: 'I
~~1. Emergency Operating Procedures I~~
EDPO, PostTrip immediate Amions l

EOP4, Functional Recovery Procedure I
~~l I j i i i l l i i i i L i i t i i r I Calvert Cirffs EAL Basis Document ago June 15.1993 j i l~~
~~.t .'~~
ne4 a 1 f t 6 i i F ELECTRICAL : I I i l l J, 1 1 l' 4 _______________._______-_8- _ _ _
i ELECTRICAL Emc_rgency Classification Level: UNUSUAL EVENT Applicable OpenrJonal Modes: ALL ' Calvert Cirffs initiatino Condition: 1 EU1 Loss of Off-Site Power NUMARC Recoonition Catego_rr System Malfunction M) MARC InitiatingDndition: SU1 Loss of All Off-Site Power to Essential Busses for Greater Than 15 Minutes Barrier Not Applicable em NUMARC Generic Basis: Pmlonged loss of AC power mduces mquimd redundancy and potentia!!y degrades the level of safety of the plant , by mndering the plant mom vulnerable to a complete Loss of AC Power (Station Blackout). Fifteen minutes was . selected as a thmshold to exclude transiert or momentary power losses. Multaanit stations with shamd safety functions should further consider how this IC may affect mom than one unit and how this may be a factor in escalating the emergency class. _P_lan_t-Specific information: Pmcedum EDP-2 Loss of Off-Site Power,would be implemented underthe conditions of concem. ADNF applies to the other operational modes when the plant is crtical. Per EON, the following am symptoms of a loss of off- i site power-
Momentary loss of Control Room lighting on both Units.

500KV Red Bus and Black Bus power available lights are de-energized.

Diesel Generators automatically start.

13KV Service Buses 12 and 22 power available lights em de-energized.

No RCPs are running on either Unit.

Reactor Trip occurs due to RCS low flow. !
For consistency with procedural mquimments and to mflect potential severity, separate EALs have been developed for hot s' d cold conditions. With the plant iaitially operating in Mode 1 or 2, EDP 2 would be entemd on a loss of off-site power. Under these conditions, restoring off-site power is expected to take no less than 15 minutes based on pmcedum implementation. Themfom, EAL 1 does not use the generic 15 minute thmshold. Thus, EAL 1 is wrtten as: l EOP-2, Loss of Off-Sita Power, implemented On Either Unit EAL 2 addmsses loss of off-site power when EDP 2 does not apply. Thus, EAL 2 is written as: Loss of Off-Site Power for GREATER THAN 15 Minutes l l Calvert Chffs EAL Basis Document E:1 June 15,1993
ELECTRICAL I l Source Documents / References / Calculations: j I
~~1. Technical Specifcations 1~~
~~TS 3.8.1 A.C. Sources~~
~~2. Emergency Operating Procedures ,~~
~~EOP-2, Loss of Offete Power '~~
~~3. Abnormal Operating Procedures~~
~~ADPGF, Loss of Off-Ste Power While in Modes 3,4,5 or 6 i~~
( Cabert Chffs EAL Basis Document E2 June 15,1993
ELEC1'RCAL Ememency Classrfication Level UNUSUAL EVENT Applicable Operational Modes: All Calvert Chffs initiatino Condtion: EU2 Loss of Vital 125 Volt DC Power for GREATER THAN 15 Minutes NUMARC Recoanidon Categor.y; System Malfunction NUMARC Initiatino Condtion: SU7 Unplanned Loss of Required DC Power During Cold Shutdown or Refueling Mode Greater Than 15 Minutes Darner: Not Applicable NUMARC Generic Basis: The purpose of this C and ts associated < Generic > EAL is to recognize a loss of DC power compmmising the abilty to monitor and control the removal of decay heat dunng Cold Shutdown or Refueling operations. This EAL is intended to be anticipatory in as much as the operedng crew may not have necessary indcation and control of equipment needed to respond to the loss. Unplanned is included in this IC and EAL to preclude the declaration of an emergency as a resut of planned maintenance acdvties. Routinely, plants will perform maintenance on a train related basis during shutdown periods. t is intended that the loss of the operating (operable) train is to be considered. If this loss resuts in the inabilty to maintain cold shutdown, the escalation to an Alert will be per <GA2, Inabilty to Maintain Plant in Cold Shutdown >. (Site-specific) bus voltage should be based on the minimum bus votage necessary for the operation of safety related equipment. This voltage value should incorporate a margin of at least 15 minutes of operation before the onset of inabilty to operate those loads. This voltage is usually near the minimum votage selected when battery sizing is performed. Typicalty, the value for the entire battery set is appmximately 105 VDC. For a 60 cell stnng of batteries the cell voltage <is typicalp 1.75 vots/ cell. For a 58 string battery set the minimum voltage is typically 1.81 vots/ cell. Plantepecife information: The 125 volt de and 120 volt vtal ac systems for the plant are d~vided r into four independent and isolated channels. Each channel consists of one battery, two battery chargers, one de bus, multiple de unit control panels, and two inverters. Each inverter has an associated vital oc distribution panel board. Power to the de bus, de unit control panels, and inverters is supplied by the station batteries and/or the battery chargers. Each battery charger is fully rated and can recharge a discharged battery while at the same time supplying the steady state power requirements of the system. A reserve 125 volt de system for the plant is completely independent and isolated fmm all four sepamtion groups, yet is capable of replacing any of the 125 vot de batteries. This system consists of one battery, one battery charger, and the associated DC sweching equipment Only the battery may be transferved for replacement duty. The 125 vot de bus 11 prtwides contml power for equipment associated with load group A for both unts. The 125 vot de bus 21 pmvides control power for equipment associated with load gmup B for both units. The 125 vot de buses 12 and 22 are used to supply power to the computer inverters, diesel generator 12 control circuts, contml room emergency lighting, and two channels of the 120 vot vtal ac system. Calvert Ckffs EAL Basis Document E3 June 15,1993 i f
B.ECTRCAL There is ons battery charger f:d from Unt 1 cnd another battary charger fed from Unt 2 connemed to each 125 vot de bus. The ac power for both banery chargers per bus is obtained from the same load group. The reserve battery is connected to its own charper when it is not connected to a safety related 125 vot de bus. Each of the four 125 vot de power sources is equipped with the following instrumentation in the contml room to enable continual operator assessment of 125 vot de power source condtion:
DC bus undervoltage a! arm

Battery current indcation

Charger current indication

Charger malfuncion alarm (including input ac undervotage, output de undervotage, and output dc ;
~~overvoltage)~~
DC bus votage indication, and

DC ground indication Components affected by the loss of 125 vot de buses 11,12,21. or 22 are listed in table EU2-1.
~~Thus, the EAL is wntten as: ADP 7J. Loss of 120 Vot Vtal AC or 125 Vot Vtal DC Power, is implemented AND 125 Vot DC Power Last for GREATER THAN 15 Minutes Source Documents /Refemnces/ Calculations:~~
~~1. Abnorrnal Operating Procedums~~
~~ADP-7J. Loss of 120 Vot Vtal AC or 125 Vot Vtal DC Power~~
2. Updated Final Safety Anatysis Report

3. BG&E Drawing 6103OE Single Une Diagram. Vtal 12DV AC & 125V DC - Emergency 250V DC

4. BG&E Drawing 61057-E Block Diagram - Auxiliary System Load Gmups - Units 1 & 2 Cahert Cirffs EAL Basis Document E-4 June 15,1993
ELECTRCAL Table E-1: Effects of Lost 125 Volt DC Buses 11,21,12, and 22 Loas of Loas ct Losa d Lnsa d 11 125 volt de Bus 21 125 vot dc Bus 12125 volt de Bus 22125 volt de Bus Channel ZD ESFAS ard AFAS Channel ZE ESFAS and AFAS Channol ZF ESFAS ard AFAS Chanrul ZG ESFAS and AFAS Sensor Catanets doenorpzed Sensor Catanots doenergized Sensor Catwnota doenorged Sensor Catuncts deonorgized CNTMT Area Red Moncor out OmAT Area Red Moretor out CNTMT Area Red Monitor out CNTMT Area Red Marrar of sorwce of service of service out of servia Channol A RPS Catanot do- Channel B APS Cotunot de- Channel C APS Catunot do- Channel D APS Catnnot do-enorged energand anorgized energized Loss cd 11 EDG fold flash and Loss of 12 EDG fidd flash and Loss of 12 EDG feid flash ard Loss of 12 EDG feld flash cortrol pomr; the start concul power, the start cortrol power, the start and control power, the start solarnds fail shut (Urvt i orty) solorods fail shut (Une 2 soionods fail shut if aligned to solurnds fail shut if algned orvy) Urvt 1 to Urst 2 , Loss of breaker posiuon Loss of treaker poscon rdcanort recataort Pbrmal power supply to the - Phrmal power supply to 11 A/21 A and 12A/22A the 11B/218 and ACPs 12B/22B ACPs 11/21,12/22.15/25,
13/23 and 14/24 4 KV and 16/26 4 KV buses buses 11 A/21 A 11B/21B, 13A/23A 138/238, 12A/22A and 12B/22B 14A/241. ard 14B/248 480 Volt Buses 480 Volt Buses 11 ard 1213 KV buses (Une 1 orty) lnse of Urut 2 Annunciauon All Urut 1 Annuncetor kghts deenorpred (Staus Panels rerneri crerpzod)
CC CNTMT SUPPLY fails shut CE CNTMT FETURN fails shut 12 SG AFW STM SUPP & 11 SG APNSTM SUPP & BYPASS vaks fail shut BYPASS vehes fad shut Loss of SRW to the Turtune (mss of SRW to the Turtune Building Building M and PA mey be lost due to M ard PA muy be lost due to loss of SRW to the Turture loss of SRW to tre Turtune Building Building Chanrel A ESFAS and AFAS Channot B ESFAS ard AFAS Actuanon Catancts de- Actuouon Catanets do. energacd enorptred 11/21 SRW,11/21 CC, and 12/22 SRW,12/22 Cn and 11/21 ECCS Purnp Roorn HK 12/22 ECCS Pump Room HX SW outlet vows fail open SW outiot ve%s fail open 11/21 Main Steam Effluom 12/22 Main Stasm Effluart Red Marutar out of sorwcc Red Moncor out of servix 11 and 12 SFP Host 11 and 12 SFP Heat Exctengers lose cookng flow Eutengers lose cookng flow duo to SRW outlet CVs faihng due to SRW outlet CVs failtrg shut (Unit 1 only) shut (Urvt 1 ordy) 11/21 MSfV loses poscon 12/22 MSIV loses poscon recauon, but mn soll be recatort but can sull te closed from 1CD3/2CO3 closed frorn 1CO3/2CD3 l l Calvert Clffs EAL Basis Document E:5 June 15,1993 1 1 l 2
ELECTRICAL Table E-1: Effects of Lost 125 Volt DC Buses ii. 21,12, and 22 _ Lose d Loso d taas d Loss d 11 125 vot de Bus 21 125 voit de Bus 12125 voit de Bus 22125 vot de Bus DJiMT Hgh Rarige Morstar DJTMT Hgh Range Moncor Denrel A out of service Dennel B out d serva loss of open signal to tre Turtmoe Bypass Va%s and loss d qud open signal to the AD/s (Unrt i oW) Aux Spray Vow fails shut IA downs:rusm of the DJTMT M Contrd Va% is isoisted [*OJTMT M 130 LATED W 2CB5CV CLDGED* alarm does PCT actuate) OITMT Gassous Moncor out of sonacc Gesnous and Uqtad Wasta - rnicase cartrol vaws fail shut (Urvt 1 ordy) 11B/218 and 128/220 RCPs are untnppstne from 1EIG/2&l6 Loss of lotdcwn due to 1/2-CVC 516CV faiing shut AFW Turtune Drwan Trun g Row Corrid Vaks 11 SG and 12 SG fail open (Une 1 ordy) POIN-404 nopoceblo n MPT DMLE [Urut 1 ordyl TCBs 1 and 5 tnp TCBs 2,6. and 9 tnp 1CBs 3 and 7 tnp TCDs 4 and B tnp On 1C13. loss of poscon On 1C13. loss of poscon rdcauors for 12 D/G SUPP recatxms for 12 D/G SUPP NO RTN LNT 2 SRW SYS NO RTN LNT 1 SRN SYS wa%s and the wa%s fail opon vtAos and the veWe fail opon Loos of r*rt osedkynph (Urut 1 o%) l Calvert Cliffs EAL Basis Document E6 June 15,1993
i ELECTRICAL - Emeroency Classifcation Lmet ALERT ,
~~.i Applicable Operational Modes: 5, 6 i Calvert Cirffs initiatino Condition:~~
EA1 Stauon Blackout While On Shutdown Coohng .: NUMARC Recoonition Cateoory System Malfunction ; NUMARC Initiatino Condition:
?
6 SA1 Loss of All Off-Site Pcwer and Loss of All DnSte AC Power to Essential Busses During Cold Shutdown i Dr Refueling Mode % Sarrier, Not Applicable I NUMARC Generic Basis: !
)
Loss of all AC power compmmises all plant safety systems requiring electric power including RHR, ECCS, l Containment Heat Removal, Spent Fuel Heat Removal and the Ultimate Heat Sink. When in cold shutdown, l mfueling, or defueled mode the event can be classified as an Alert because of the significantly reduced decay. ! heat, irrwer temperature and pressum, increasing the time to mstore one of the emergency busses, relative to ; that specified for the Site Emergency EAL Escalating to Site <>mergency, if appmpriate, is by < Radioactivity i Release or SEC> Judgement ICs. Rfteen minutes was selected as a threshold to exclude transient or momentary ! power losses. Plant 4peerfic information: ! Section IX of ADPGB addmsses loss of 4 kV power suppies interrupting shutdown cooling. This EAL addresses : Statson Blackout conditions during cold shutdown or refueling, Thus, the EAL is wrc.en as: , ADP 38 Abnormal Shutdown Cooling. is implemented Due to Loss of 4 kV Power Supplies For GREATER THAN 15 Minutes Source Documents / References /Celculations:
~~1. Abnormal Operating Procedures ,~~
~~ADP-38 Abnormal Shutdown Cooling l~~
l l l l Cahert C!rffs EAL Basis Document E:7 June 1C.1993 i i
ELECTT1 CAL I l Ememency Classifcation Le_ve_l: ALERT s Applicable D;e_ rational Modes: 1,2,3,4 Cakert cirffs initiatino condition: EA2 Only One AC Power Soume Available to Supply 4 kV Ememency Buses NUMARC Recoonition CateJo_ry System Malfunction NUMARC initiatino Condtion:
SAS AC Power Capabilty to Essential Busses Reduced to a Single Power Source for Greater Than 15 Minutes Such That Any Addtional Single Failure Would Result in Station Blackout Barner. Not Appleable NUMARC Generic Basis:
This C and its associated < Generic > EAL are intended to provide an escalation fmm C <EU1, Loss of Offete Power >. The condtion indicated by this C is the degradation of the offsite and on-site power systems such that any addtaonal single failure would result in a station blackout. This condition could occur due to a loss of off+ite power wth a concurrent failure of one diesel generator to suppy power to its emergency busses. Another related condtion could be the loss of all off-site power and loss of on-sce emergency diesels with only one train of emergency busses being backfed fmm the unit main generator, or the loss of on-site emergency diesels with ony one train of emergency busses being backfed from offsite power. The subsequent loss of this single power source would escalate the event to a See <E>mergency in accortience with C <ES1, Station Blackout >.
<Genenc> EAL ib should be expanded to identify the control room indcations of the status of Site-specific power sources and distribution busses that, if unavailable, establish single failure vulnerabilty.
At multionit stations, the EALs should allow credt for operation of installed design features, such as cmss-ties or swing diesels, provided that abnormal or emergency operating procedures address their use. However, these stations must also consider the impact of this condtion on other shared safety functions in developing the site specific EAL Plant-Specife information: The EAL addresses condtions while operating in Modes 1,2,3, or 4 under which only one method is available to supply the emergency buses and loss of that method will result in a Station Blackout. Thus, the EAL is wntten as: Ony One Source (Off-Site or Desel) Available to Supply Bus 11 Dr 14 (Bus 21 Dr 24) for GREATER THAN 15 Minutes AND Unit Not on Shutdown Cooling Under conditions where diesel generator 12 is suppying poser to one Unit, it should not be considered available as a power source for the other Unit. '
~~/ )~~
Calvert Cliffs EAL Basis Document E8 June 15,1993 i
ELECmCAL Source Documents / References /Calculationg-(' ' \
~~1. Updated Final Safety Analysis Report~~
~~Section 8, Electric Power Systerns~~
~~2. Emergency Operating Procedures~~
~~EOP2 Loss of Off6ite Power~~
~~3. Technical Specifications~~
~~TS 3.8.2. Or>Srt.e Power Distribution Systems I~~
(~) L : i O Cabert Ctrffs EAL Basis Document ES June 15,1993 ; l
ELECTRCAL Em.ftrgency Classification Level: ALERT Applicable Operational Modes: 1, 2 Qdart_Cirffs initiating Condition: EAS Loss of 125 Volt DC Power AND Reactor Trip NUMARC Recognition Ca_tegory; System Malfunction NUMARC Initiatino Condition: SS3 Loss of All Vital DC Power Barrier: Not Appicable NUMARC Generic Basis-Loss of all DC power compromises abihty to monitor and control plant safety functions. Prolonged loss of all DC power will cause core uncovering and loss of containment integrity when there is significant decay heat and sensible heat in the reactor system. Escalation to a < Site > Emergency would occur by < Radioactivity Release. Fission Product Barrier Degradation, Equipment Failure, or SEC Judgement Cs> Fifteen minutes was selected as a threshold to exclude transient or momentary power losses. Mu!*J unit stations with shared safety functions should further consider how this O may affect more than one unit and how this may be a faaor in escalating the emergency class. Plant-SpecWie Information: s The Vital (Class 1E) 125 V DC power system is fully described under C EU2, Loss of Vital 125 Volt DC Power for GREATER THAN 15 Minutes. Review of the information in Table E-1 shows that if either DC bus 11 or 21 l were lost with at least one unit in operation the resulting plant response meets the threshold for an Alert at Catvert Cliffs. The EAL is writzen to be consistent with procedures applying to plant operation while the reactor ! is critical Thus, the EAL is wrcten as: EDPB, Functional Recovery Pmcedure, is implemented on Loss of 125 Volt DC Bus _ Source Documents / References / Calculations:
~~1. Abnorvnal Operating Pmcedures~~
~~ADP-7J. Loss of 120 Volt Veal AC or 125 Volt Vrtal DC Power~~
~~2. Emergency Operating Pmcedures~~
~~EOPE, Functional Recovery Procedure~~
3. Updated Final Safety Analysis Report

4. BG&E Drawing 61030 E. Sing?e Line Diagram Vital 12DV AC & 125V DC - Emergency 25DV DC

5. BG&E Drawing B1057f., Block Diagram - Auxiliary System Load Gmups - Units 1 & 2 Calvert Chffs EAL Basis Document E1D June 15,1993
~~. -. - _ - . . - --- .- ~~ .- . . - . .=- - - -~~
t EECTRICAL i Emergency Classifcation Levet STE EMERGENCY l Applcable Doerational Modes: .1,2,3,4 Cakert Cliffs initiatino Condition: ES1 Atatir=1 Blackout i NUMARC Recoonition Cateoory System Malfunction NUMARC initiatino Condition: :? SS1 Loss of All Offsite Power and Loss of All Onsite AC Power to Essential Busses i Barrier: Not Applicable NUMARC Generic Basis: Loss of all AC power compromises all plant safety systems mquiring electre power including RHR, ECCS, Containment Heat Removal and the Ultimate Heat Sink. Prolonged loss of all AC power will cause core uncovering and loss of containment integrity, thus this event can escalate to a General Emergency. The (Staspecife) time duration should be selected to exclude transient or momentary power losses, but should not exceed 15 minutes. Escalation to General Emergency is via Fission Product Barrier Degradaten or C <EG1 Prolonged Staten ; Blackout >. i MultkJnit stations with shamd safety functions should further consider how this IC may affect mom than one unit .f and how this may be a factor in escalating the emergency class. Bant-Specife Infomnation: The Calvert Cirffs EAL is based on NUMARC Entry into EDP 7 cortssponds to the NUMARC specified conditions. f Under these conditions, it is expected that restoring off-site power would take greater than 15 minutes. : 1 Therefore, CCNPP does not include the generic 154ninute threshold. Thus, the EAL is wntten as: . EOP-7. Staten Blackout, is implemented i' Source Documents / References / Calculations:
~~1. Technical Specircations~~
~~TS 3B.1. A.C. Sources !~~
~~2. Emergency Operating Pmeedures~~
~~EDP-7. Station Blackout ~~~
i
~~3. Updated Final Safety Analysis Report~~
~~Section B. Electre Power Systems Cabert Cirffs EAL Basis Document June 15,1993 -f E11~~
~~~ M Ememency Classification Levet STTE EMERGENCY (f Applicable Operational Modes: All Cabert Cliffs initiatino Condttion:~~
ES2 Loss of All 125 Volt DC Buses NUMARC Recoonition Cateoorv: System Malfunction NUMARC initiatino Condcion: SS3 Loss of All Vital DC Power SS6 Inabilcy to Monitor a Signifcant Transient in Pmgress Barries Not Appleable NUMARC Generic Basis: [SS3] Loss of all DC power compmmises abilty to monitor and contml plant safety functions. Pmlonged bss of all DC power will cause com uncovering and bss of containment integrity when there is significant decay heat and sensible heat in the reactor system. Escalation to a General Emergency would occur by < Radioactivity Release. Fission Product Banier Degradation, or SEC> Judgement Cs. Rfteen minutes was selected as a threshold to exclude transient or momentary power bsses. Multkinit stations with shamd safety functions should further consider how this IC may affect more than one unit C and how this may be a factor in escalating the emergency class. [SS6] This IC and its associated < Generic > EAL am intended to recognize the inabilty of the contml room staff to monitor the plant response to a transient. A See <E>mergency is considered to exist if the contml mom staff cannot monitor safety functions needed for the protectbn of the public. (Sce-specific) annunciators for this EAL should be limited to include those identified in the Abnonnal Operating Procedures, in the Emergency Operating Procedums, and in other EAls (e.g., red monistors, etc.). Compensatory non-alarming indications in this context include computer based information such as SPDS. This should include all computer systems available for this use depending on specifc plant design and subsequent retmfes. Sgnificant transient includes msponse to automatic or manually initiated functions such as scrams, runbacks involving greater than 25% thermal power change, ECCS injections, or thermal power oscillations of 10% or greater, t (Sce Specific) indcations needed to monitor safety functions necessary for pmtectbn cd t% public must include control mom indications. computer generated indcations and dedicated annunciation capabilty. The specific indcations should be those used to determine such functions as the abilty to shut down the reactor, maintain the core cooled and in a coolable geometry, to remove heat fmm the cors, to maintain the reactor coolant system intact, and to maintain the containment intact. Planned actions are excluded fmm this EAL since the loss of instrumentation of this magnitude is of such signifcance dunng a transient that the cause of the bss is not an emeliorating factor. Calvert Cliffs EAL Basis Document E12 June 15.1993
~~. - - - . _ - _ _ _ _ a __~~
GECTP'*N i Bant6pecific information: Because of the 125 Vot DC and Annunciator design at Calvert Cliffs. NUMARC Cs SS3 and SS6 have been ( combined into one C for Cabert Cliffs. The Vtal (Class 1E) 125 V DC power system is fully described under C EU2. Loss of Vcal 125 Volt DC Power for GREATER THAN 15 Minutes. Review of the information in Table E-1 - shows that if all 125 Vot DC buses were lost, the resuting plant response meets the threshold for a Site i Emergency. Thus. the EAL is wrcen as: Loss of 125 Vot DC Buses 11,12,21 And 22 , Source DocumentsjReferences/ Calculations:
~~1. Abnormal Operating Procedures~~
~~ADP-7J. Loss cf 120 Vot Vtal AC or 125 Volt Vt&tCPower~~
~~2. Emergency Operating Procedures~~
~~EOPO, PostrTrip Immediate Actions '~~
~~3. Updated Final Safety Analysis Report~~
~~Section B, Electric Power Systems l~~
l l l 4 l Cabert Cidfs EAL Basis Document E:13 June 15,1993 l l
i ELECTRCAL > Ememency Cla_nsifcation Level: GEMRAL EMERCENCY Applicable Doerationaf Modes: 1,2,3,4 Calvert Cidfs initiatino Conditio_g: i EG1 P, Aged Sation Blackout f N1JMARC Recoonition Cateoo_ry; System Malfunction j NUMARC Initiatino Conditjo : SG1 Prolonged Loss of All Off4ce Power and Prelonged Loss of All Dnsite AC Power Barrier: Not Applicable t , NLUMARC Generic Basis: Loss of all AC power compromises all plant safety systems requiring electre power including RHR, ECCS, Containment Heat Removal and the Utimate Heat Sink. Pmlonged loss of all AC power will lead to loss of fuel + clad, RCS, and containment. The (Site-specife) hours to restore AC power can be based on a site blackout coping analysis performed in conformance with 10 CFR 50.83 and Regulatory Guide 1,155, Staten Blackout, as available, with appmpriate allowance for offsita emergency response. Although this C may be viewed as redundant to the Fission Produ1 Bamer Degradation C, its inclusion is necessary to betrar assure timely recogntion and emergency msponse. This C is specified to assure that in the unlikely event of a pmlonged station blackout, timely recognition of the , seriousness of the event occurs and that declaration of a General Emergency occurs as earty as is apprepriate, based on a masonable assessment of the event trajectory
~~<The likelihood of restoring at least one emergency bus a,hould be based on a resiste appresol of the aben [~~
i since a deley in an upgrade decision based on only a chance of mitigating the event could result in a loss of 4 veluable time in preparing and implementing public protective artmns> in adddon, under these condcions, fission product barrier monitoring capabilty may be degraded. Although it may be ddfeult to predict when power can be restored, it is necessary to give the <SEC> a reasonable idea of how quickfy (s)he may need to declare a General Emergency based on two major considerations:
1. Are there any present indications that core cooling is already degraded to the point that Loss or Potential Loss of Fission Pmduct Barriers is MMINEN77 < Refer to Fssion Product Barrier Degradation EAL Table for more information>.

2. If there are no present indications of such core cooling degradaton, how likely is it that power can be restored in time to assure that a loss of two barriers with a potential ioss of the third barrier can be prevented? i Thus,indcation of continuing core cooling degradation must be based on Fission Preduct Barrier monitoring with partcular emphasis on <SEC> judgement as it relates to MMINEhTT Loss or Potential Loss of fssion product barriers and degraded abilty to monitor fission preduct barriers.
Multionic stations with shared safety functions should further consider how this C may affect more than one unit and how this may be a factor in escalating the emergency class. Cafvert Cirffs EAL Basis Document E:14 June 15,1993
_ ELECTRICAL Plant Soecific information: Under conditions whem e diesel generator is suppying power to one IJbit, it should not be considered available ) es a power supply for the other Unit. The first part of this EAL corvesponds to the threshold conditions for IC ES1, Station Blackout for GREATER THAN 15 Minutes. The second part of the EAL addresses the conddons that will escalate the SBO to General Emergency. Occurrence of any one of these conddons following SBO is sufficient for escalation to General Emergency. These conditions are: (1) SBO coping capability, or [2] indications of inadequate core cooling. Each of these conddons is discussed below-
1. SSO Cooino Cacability Calvert Cirffs fe!!s wrthin the four hour SBO coping category. The abilty of each unit to cope with a four hour SBD duration was based on an assessment of condensa:a inventory requimd for decay heat mmoval. Class 1E bat:ery capacity, compressed air availabilty or manual operetion of certain valves, effects of loss of ventilation, containment iaolathn vatse operability, and reactor coolant inventory loss. A plantespectic analysis indicates that the expecteo'iates of reactor coolant inventory loss under SBO conddons do not result in com uncovery in a SBO of four hours. Therefore, makeup systems in addtaon to those curmrch available under SBO conditions am not requimd to maintain com cooling under naturel circulation [ including mflux boiling). Thus, conditions in which restoration of AC power within 4 hours is NOT likely am included in the EAL

2. 1n_dications of inadeauste Core Coolino Calvert Chffs does not use Cntscal Safety Function Status Trees. Calvert Cliffs uses Safety Function Status Checks developed by the Combustion Engineering Owners' Group (GE OG) which are based on logic similar to that used for CSFSTs developed for Westinghouse PWRs The applicable acceptance creeria for Com and RCS Heat Removal are shown on the Safety Function Status Checks. They are:
~~S:aam Generators Available for RCS Heat Ramcual~~
1. Adequate secondary side liquid inventory in at least one steam generator as indicated by level between -170 and +30 inches, and

2. Adequate source of feedwater available to assure continued liquid inventory available as indicated by Condensate Storage Tank level greater than 5 feet, and

3. Steam Generators actirig as effeczke heat sink as indicated by Cold Leg Temperatums (Ty constant or lowering.
~~Pnmary Side Conddons for Core and RCS Hast Ramcual~~
1. Adequate core heat removal as indicated by Com Exit Thermocouple readings less than superheated, and

2. Ether of the following:
~~Natural circulation established as indicated bythe temperature difference between Hot Leg Temperature~~
~~[T,) and Tm of between 10 *F and 50 *F, or~~
~~Forred circulation effective as indicated by T,- Tm less than 10 *F.~~
Per CEN-152, supertiested condtions indicate core uncovery and inadequate core cooling. O a Catvert Chffs EAL Basis Document E:15 June 15,1993
l EMCA'_ l Thus, the EAL is wrnen as: 1 C T EOP-7, Station Blackout, is implemented AND Any of the Following:
Restoration of Power to ANY Vtal 4kV Bus is NOT Ukety Wthin 4 Hours e Valid CET Readings indicate Superheat Temperatums

Core and RCS Heat Removal Using Steam Generators Can NOT Meet Acceptance Cntena Valid means that the indication is from instrumentation determined to be operable in accordance with the Technical Specications or has been venfied by other indications displayed on the contml panels. ;
Can NOTis used because the ability to meet the final acceptance criteria is the appropriate concem, not whether intermedae acceptance creeria are not being sch'eved at any given moment. f Source Documents /Refemnces/ Calculations:
~~1. Emergency Operating Pmcedums~~
EDP-7 Station Blackout

EOPa. Functional Recovery Procedum
2. CEN-152. Emergency Pmcedum Guidelines

3. Let:er. Daniel G. MacDonald (US Nuclear Regulatory Commission) to G.C. Creel (BGSE), Response to Ekation Blackout Rule - Ca! vert Cirtfs Nuclear Power Plant, Unts 1 and 2, TAC Numbers 68525 (Unit 1) and 68256 (Unit 2), October 10,199D N
[ \ Caf/ert Clrffs EAL Basis Document E:16 June 15.1993 1
b b i t L i I t h
~~'l t~~
~~( e p h p~~
?
P SECURITY r I I l 1 I l r s I 1 i l I i 1 i l l
I SECURTIY . Egargency Classifcadon Level: UNUSUAL EVENT Applicable Operational Modes: ALL _Cabert Cidfs initiadng Condition: TU1 Confirmed Secunty Event With Potantaal Degradation in the Level of Safety of the Plant NUMARC Recoonition CateD91y Hazards and Other Condtions Affecting Plant Safety BlUMARC intiatino Condtion: HU4 Confirmed Secunty Event Which Indicates a Potential Degradation in the Level of Safety of the Plant Not Applicable 13arrier: NUMARC Generic _L3 asis: This EAL is based on (Ste-spectc) See Security Plan. Secunty events which do not mpresent at least a potential degradation in the level of safety of the plant are reported under 10 CFR 73.71 or in some cases under 10 CFR 50.72. The plant Protected Area Boundary is typicaly that part within the secunty isolation zone and is defined in the (Ste-specife) security plan. Bomb devices discovemd within the plant Vtal Area would result in o escalation 9.o a higher emergency classifcation level via other Security Event ICs>. Plant @pecife information: The Calvert Cidfs EALs address the genere amas of concem and include the ISFSt. Thus, EAL 1 is wrcten as: I
*Secunty Emergency" or "Secunty Alert' Declared for:
~~Sabotage Wsthin or to Plant Prutected Area e intrusion into or Attempted intrusion in Plant Protected Area EAL 2 is wntten as:~~
~~"Secunty Event" Declared for.~~
~~Sabotage Wehin or to ISFSI Protected Area e Perimeter Intrusion into ISFSI Protected Area Soume Documents / References / Calculations:~~
~~1. BG&E Intemal Memorandum, Tom Forgette (Emergency Planning Unit) to POSRC, Juy 29.1986 l~~
' l l 1 Calvert Cliffs EAL Basis Document June 15,1993 l T:1 l l t
i SECURITY Emergency Classification Levet: ALERT Apphcable Doerational Modes: ALL Cakert Chffs initiatino Condition: TA1 heity Event in the Rant W Area NUMARC Recoonition Catecory Hazards and Other Conditions Affecting Plant Safety , NUMARC initiatino Conditign: HA4 Security Event in a Plant Protected Area Barrier: Not Applicable NUMARC Generic Basis: This class of security events mpresents an escalated threat to plant safety above that contained in the Unusual Event. For the purposes of this IC, a civil disturbance which penetrates the pmtected area boundary can be considered a hostile forte. Intrusion into a vital area by a hostile force will escalate this event to a Site
~~< Emergency >.~~
Plant 40ecific information: The Catvert Cliffs EALs address the generic areas of concem. Thus. EAL 1 is wntzen as: Forced Entry of Unauthonzed Personnel into the Vrtal Area Affecting the Aoility to Acheve or Maintain ( Safe Shutdown EAL 2 is wntten as: Sabotage of Vrtal Area Equipment in Progress Affecting the Abiltty to Achieve or Maintain Safe Shutdown l The list of areas of concem for Safe Shutdown are shown below and are prominently displayed on the EAL Table. i i Calvert Chffs EAL Basis Document T:2 June 15,1993
SECURITY Arsas of Concem for Safe Shutdown
Contml Room

Electrical Penetration Rooms

Control Room HVAC Room

Auxilary Feedwater Pump Room

Cable Spreading Room

Chaming Pump Rooms

Cable Chases

Desel Generator Rooms

bitchgear Room

Refueling Water Tank (RWT) 11(21)

ECCS Pump Room

Condensata Storage Tank (CST) 12

Service Water Pump Room

Pmtmated Water Storage Tank (PWST) 11(21) )

Component Cooling Pump Room

Fuel Oil Storage Tank (FDST) 12 ;

Main Steam Penetration Room This list of Safe Shutdown amas is disp',Erged on the EAL Tables to assure that all areas related to Safe Shutdown are considered by the SEC.
~~Source Documents / References / Calculations:~~
~~1. BG&E Intemal Memorandum, Tom Forgetta (Emergency Planning Unit) to POSRC, July 29,1988 I~~
J l O i \ Calvert Cliffs EAL Basis Document T;3 June 15,1993
SECURITY Ememency Classification Levet SITE EMER37EY Acolicable Operational Modes: ALL Cakert Cliffs initiatino Condition: TS1 Security Event in a Plant Vital Area NUMARC Recoonition Cateoore Hazards and Other Condtions Affecting Rant Safety NUMARC Initiatino Conditior}: HS1 Security Event in Rant Vcal Area Barrier: Not Applicable NUMARC Generic Basis: This class of security events represents an escalated threat to plant safety above that contained in the Alert IC in that a hostile force has progressed fmm the Protected Area to the Vtal Area. < > Rant @pecific Information: The Cakert Cliffs EALs address the generic areas of concem. Thus, the EAL is wntten as: Security Threat Resuting in imminent Loss of Abilty to Achieve And Maintain Safe Shutdown of Either Reactor The list of areas of concem for Safe Shutdown are shown below and are prominenth displayed on the EAL Table. Areas of Conosm for Safe Shuttbwn
Contml Room

Electrical Penetraton Rooms

Control Room HVAC Room

Auxiliary Feedwater Pump Room

Cable Spreading Room

Charging Pump Rooms

Cable Chases

Diesel Generator Rooms

Switchgear Room o Refueling Water Tank (RWT) 11(21)

ECCS Pump Room

Condensate Storage Tank (CST) 12

Service Water Pump Room

Pretreated Water Storage Tank (PWST) 11(21)

Component Cooling Pump Room

Fuel Oil Storage Tank (FOST) 12

Main Steam Penetration Room This list of Safe Shutdown areas is displayed on the EAL Tables to assure that all areas related to Safe Shutdown are considered by tre SEC.
~~Source Documents / References / Calculations:~~
~~1. BG&E Intemal Memorandum, Tom Forgette (Emergency Ranning Unt) to POSRC, Juy 29,198S V~~
Cakert Cidfs EAL Basis Document T:4 June 1S,1993
SECURmf Ememency Classification Level: GEhERAL EMERGENCY Acoticable Doeretional Modes: All Cabrert Cliffs initiatino Conddon: TG1 Security Event Resulting in Loss of Abifity to Reach AND Maintain Cold Shutdown NUMARC Recoonition CatecoJy Hazartis and Other Condbons Affecting Plant Safety NLJMARC Initiatino Condition: HG1 Security Event Resulting in Loss of Ability to Reach and Maintain Cold Shutdown Barrier: Not Applicable NUMARC Generic Basis: This C encompasses conditions under which a hostile forte has taken physical control of vital ama required to reach and maintain safe shutdown. < > Bant@gecific nt Information: The Cafvert Cliffs EALs addmss the generic areas of concem. Thus, the EAL is wroten as: Security Threat Resulting in Loss of Ability to Achieve and Maintain Safe Shutdown of Either Reactor This would include areas whem any switches that transfer control of safe shutdown equipment to outside the control room am located. The list of areas of concem for Safe Shutdown are shown below and are prominently displayed on the EAL Table. Areas of Concem for Safe Shutdown
Control Room

Electrical Penetraton Rooms

Cort.rol Room HVAC Room

Auxiliary Feedwater Pump Room

Cable Spreading Room

Charging Pump Rooms i

Cable Chases

Diesel Generator Rooms )

Switchgear Room

Refueling Water Tank (RWT) 11(21) )

ECCS Pump Room

Condensata Storage Tank (CST) 12

Service Water Pump Room o Pmtreated Water Storage Tank (PNST) 11(21) 4

Component Cooling Pump Room

Fuel Dil Storsge Tank (FOST) 12 )

Main Steam Penetration Room This list of Safe Shutdown areas is displayed on the EAL Tables to assum trat all areas related to Safe ;
~~Shutdown am considered by the SEC. I Source Documents / References / Calculations: 1~~
~~1. BG&E Intemal Memorandum, Tom Forgette (Emergency Planning Unit) to POSRC, July 29.1986 l~~
Cahtert C!rffs EAL Basis Document T5 June 15,1993 l
1 I l 1 i i l 1 1
~~.I l~~
l l l l 1 J i l l 1 i l FIRE f i 4 h l i i 4 i I l 1 1 i l l l l l l 1 I
e FIFE Emg_rgency Classification Levet UNUSUAL EVEf6 o applicable Operational Modes: A!.L Catvert Cliffs initiatino Condition: IU1 Fre Within Protartad Area Boundary Not Extinguished Wahin 15 Minutas of Detection - NUMARC Recognition Catego!y Hazards and Other Conditions Affecting Plant Safety NUMARC Initiatina Condition: i HU2 Fire Within Pmtected Area Boundary Not Extinguished Wehin 15 Minutes of Detection
~~-i Sarrier: Not Applicable NUMARC Generic Basis:~~
The purpose of this C is to address the magnitude and extent of fires that may be potentially significant . precursors to damage to safety systems. This excludes such items as firss within administration buildings, waste- ! basket fires, and other small fires of no safety consequence. This C applies to buildings and areas contiguous to plant vital areas or other significant buildings or areas. The intent of this C is not to include buildings (Ls., wamhouses) or amas that are not contiguous or immediately edacent i to plant vital amas. Verification of the alarrn in this context means those actions taken in the contml mom to determined that the control room alarm' l is not spurious. ! Escalation to a higher emergency class is by C <lA1, Fim or Explosion Affecting Safe Shutdown >. <> ' Plant-Specific Information: Each Calvert Cliffs unit uses the Abnormal Operating Pmcedures (ADP) 9A through 9S to address fires within the plant protected and vital areas that are of particular concem because they contain equipment required for safe shutdown. Fire in the Control Room HVAC Room may lead to power being lost to the altamate shutdown panels. Thus, the Contml Room HVAC Room (Room 512) has been added to the areas of concem for safe shutdown. Armas of Concem for Safe Shutdown
Control Room

Electrical Penettstaon Rooms

Control Room HVAC Room

Auxiliary Feedwater Pump Room

Cable Spreading Room

Charging Pump Rooms

Cable Chases

Diesel Generator Rooms

Switchgear Room

Refueling Water Tank (RWT) 11(21)

ECCS Pump Room

Condensate Storage Tank (CST) 12

Service Water Pump Room

Pretreated Water Storage Tank (PWST) 11(21)

Component Cooling Pump Room

Fuel Oil Storage Tank (FOST) 12

Main Steam Penetration Room This list of Safe Shutdown areas is displayed on the EAL Tables to assure that all areas related to Safe Shutdown are considered by the SEC.
~~Calvert Cirffs EAL Basis Document t1 June 15,1993 I~~
~~.n - - - ,~~
F1h - Thus, the EAL is written as: b V Fire Within An Ama Containing Safe Shutdown Equipment Lasting GREATER THAN 15 Minutes Soume Documents /Refemnces/ Calculations:
~~1. Abnormal Operating Procedures .~~
~~ADP-9A thmugh SS Altemate Safe Shutdown /Contml Room Evacuaton pmcedum series~~
~~2. Issue Report IRDO12603, Fire in Room 512 .,10-23-92 i~~
O O Cakert Cidfs EAL Basis Document 12 June 15,1993
FIRE Ememency Classrfication Levet ALERT Applicable Doerotional Modes: ALL CaVert Qltffs initiatino Condition: lA1 Fire or Explosion Affecting Safe Shutdown NUMARC Rscoonition Gat.ggog Hazards and Other Condtions Affecting Rant Safety NUMARC initiatino Condition: HA2 Fire or Explosion Affecting the Operabilty of Plant Safety Systems Required to Establish or Maintain Safe Shutdown Elarrier: Not Applicable NUMARC Generic Basis: (Site-specific) Areas containh.g functions and systems mqdred for the safe shutdown of the plant should be specified. (Site-Specific) Safe Shutdown Analysis should be consulted for equipment and plant areas required for the applicaole mode. This will make it rssier to determine if the fire or explosion is potentially affecting one or more trains of safety systems. Esco.ation to a higher emergency class, if appmpriate, will be based on
~~< Equipment Failure. Electrical. Fission Pmdua Barrier Degradation. Radoacivity Release, or SEC Judgement ICs>.~~
With regard to explosions, only those explosions of sufficient fome to damage permanent structures or equipr :w?- required for safe operation within the identified plant area should be considered. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastrophic failure of pressurized equipment, that potentially imparts significant energy to nearsby structures and materials. The inclusion of a " report of visible damage" should not be interpreted as mandating a lengthy damage assessment before classification. No attempt is made in this
< Generic > EAL to assess the actual magnitude of the damage. The occurrence of the explosion with reports of evidence of damage (e.g., deformation, scorching) is sufficient for the declaration. The declaration of an Alert and the activation of the TSC will pmvide the <SEC> with the resounces needed to perform these damage assessments. The <SEC> also needs to consider any securcy aspecs of the explosions, if applicable.
Plant-Sp_ecific Information: Each Calvert Cirffs unit uses the Abnormal Operating Pmcedures (AOP) 9A through SS to address fires within the plant protected and vital areas that are of particular concem because they contain equipment mquired for safe shutdown. 1 Thus, the EAL is wrtten as: Fim or Explosion Affecting the Ability to Achieve Or Maintain Safe Shutdown Determination of whether the fire is affecting ability to achieve or maintain safe shutdown is determined by physical observation, or by Contml Room / local control station indicatioris. Damage to one train of safe shutdown equipment when other redundant equipment / trains are operable does not affect the abilty to achieve or maintain safe shutdown (e.g., damage to an auxiliary feedwater pump when at least one other auxiliary feedwater pump is operable). Calvert C!rffs EAL Basis Document L3 June 15.1993
RAE Fim in the Control Room HVAC Room may lead to power being lost to the altamate shutdown panels. Thus, the Control Room HVAC Room (Room 512) has been added to the areas of concem for safe shutdown The list of l areas of concem for Safe Shutdown am shown below and are prominently displayed on the EAL Table. Areas of Concem for Safe ShLtdown
Control Room

Electncal Penetraton Rooms l

Contml Room HVAC Room

Atailiary Feedwater Pump Room :

Cable Spreading Room e Charging Pump Rooms ;

Cable Chases

Diesel Generator Rooms '

Swechgear Room

Refueling Water Tank (RWD 11(21) :

ECCS Pump Room

Condensate Storage Tank (CST) 12 i

Service Water Pump Room

Pretmated Water Storage Tank (PWST) 11(21) ;

Component Cooling Pump Room o Fuel Oil Storage Tank (FOST) 12

Main Steam Penetration Room This list of Safe Shutdown areas is displayed on the EAL Tables to assure that all areas related to Safe I Shutdown are considered by the SEC.
~~Source Documents / References / Calculations: ;~~
~~1. Abnormal Operating Procedures ;~~
~~ADP-9A through 9S, Atemate Safe Shutdown /Contml Room Evacuation procedure series~~
~~2. Issue Report IROO12603, Fire in Room 512 ,10-23-92 f~~
i
~~-l i~~
i
~~-l 1~~
i 1 i a h Calvert Cliffs EAL Basis Document t4 June 15,1993 i
O l l l l l l l l l NATURAL PHENOMENA O O
NARJRAL PHEh0MENA Emerne_n. Ley _Classifcation Levet UNUSUAL EVENT f" ( _AppJeable Djerational Modes: ALL \ Catvert Cliffs initiatino Condttion: NU1 Natural Phenornena NUMARC Recoonition Cateoorv. Hazartis and Other Condmions Affecting Plant Safety NUMARC Initiatino Condmion: HU1 Natural and Destructive Phenomena Affecting the Pmtected Area Barrier: Not Applicable NUMARC Generic Basis: The protected sma boundary is typically that part within the security isolation zone and is defined in the sita secunty plan.
< Generic > EAL 1 should be developed on Site-Specific basis. Damage may be caused to some portions of the site, but should not affee abilcy of safety functions to operate. Method of detection can be based on instrumentation, validated by a reliable source, or operstor assessment. As defined in the EPRkponsored Guidelines for Nuclear Plant Response to an Earthquake, dated October 1989, a " felt earthquake
~~is:~~
An earthquake of sufficient intensity such that- (a) the vibratory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of control room operators on duty at the time, and (b) for plants with operable seismic instrumentation,the seismic swtches of the plant am activated. For most plants with seismic instrumentation, the seismic switches are set at an acceleration of about DDig.
< Generic > EAL 2 is based on the assumption that a tomado striking (touching down) within the protected boundary may have potentially damaged plant structures containing functions or systems required for safe shutdown of the plant. If such damage is confirmed visually or by other irmlant indications, the event may be escalated to Alert. <3eneric> EAL 3 allows for the control room to determine that an event has occurved and take appropriate action based on personal assessment as opposed to venfication (ie., an earthquake is felt but does not register l on any plant 4pecific instrumentation, etc.) '
~~Generic EALs 4, 5, and 6 are addmssed under C OU3, Destructive Phenomena.~~
< Generic > EAL 7 covers other (Sce Specific) phenomena such as hurricane, flood, or seiche. These EALs can also be precursors of more serious events, in particular, sites subjectto severs weathar (as defined in NUMARC station blackout initiatives) should include an EAL based on salvataon of the severe weather mitigation procedures (e.g., precautionary shutdowns, diesel testing, staff callouts, etc.) < >
ElantcSoecife Information: Catvert Cliffs EALs are based directly on NUMARC and include the Independent Spent Fuel Storage installation (ISFSI) O I Calvert Ctrffs EAL Basis Document N:1 June 15,1993 j 1
NATURAL PHENOMENA EAL 1 addresses seismic activity. This EAL is based on the acceleration level which causes actuation of the seismic monitor and is verified to be the result of an earthquake, or the presence of a " felt earthquake" as descnbed in the NUMARC generic basis above. On the basis of applicable plant pmcedums, EAL 1 is written as: Earthquake Detected By Seismic Instrumentation per DM6 Or Based on Shift Supervisor Judgement EAL 2 is wrcten as: Nuclear Secunty Report of a Tomado Striking Switchyard, Plant Pmtected Ama Or Within ISFSI Protected Area Per UFSAR Section 2.8.3.4, the design basis hurricane (used for tidal surge estimates) has a maximum wind speed of 124.7 MPH and a forward speed of 23 MPH. EAL 3 uses 75 MPH to be anticipatory of the design basis wind speed. Thus, EAL 3 is wntten as: Sustained Wind Speed GREATER THAN 75 MPH (34 Meters /Second) for AT LEAST 15 Minutes The duration of 15 minutes is selected to indicate sustained winds and to pmclude wind gusts. An incmase in sustained speed above 90 mph (40 meters /second)is cause for escalation to en Alert. Wind speeds are also pmvided here in meters /second for dose assessment input. The conversion equation is as follows: m es meer me ers 75 =528 d = = =34 hour mile 3600 seconds 32808 feet second Per UFSAR Section 2.8.3.6, the still water levet used for intake Structum analysis is 17.6 feet MSL This is above the top of the range of the Tide Level Rvorder (OtR-5195]. The top of the intake Structure flood lights (located on the east side to the traveling somenslis 15 to 16 feet MSL EAL 4 is anticipatory of the design water level. Thus EAL 4 is wrcten as: Bay Water Level Above the Top of the intake Structure Rood Lights On East Side of Traveling Semens Per UFSAR Section 2.8.3.7, the predicted extmme low tide is -3.6 feet MSL and normal operation can continue i with the bay level as low as 4.0 feet MSL j l Thus, EAL 5 is written as: Bay Water Level is AT LEAST 3.6 Feet Below Mean Sea Level as indicated by OLR-5195 Tide Level Recorder Or by Physical Measurement Physical measurements may be required if the Tide Level recorder is out of service. For example, the Surveillance Test Pmcedums prcvide a way to determine Bay level using a mpe. Source Documents /Refemnces/ Calculations:
1. Updated Final Safety Anatysis Report

2. Operating instruction (OI) 46, Seismic Measurement Equipment 1
Celvert Cirffs EAL Basis Document N:2 June 15,1993 l
NATURAL PHENOMENA
3. BGE Drawing 60 220E (M-31), Equipment Location Service Building, Water Tmatment Ares & Intake Structure Section "J J'

4. BGE Drawing 83-278-E, Plan Auxiliary Building Restricted Access Ama El (-)B'O*,(-11D'O' And (-115'O'

5. BGE Intemal Memorandum, JE Thorp to RE Denton, Emergency Action Level Review Criteria, June 1, 1990

6. Letter, G.C. Cmel (BGM) to U.S. Nuclear Regulatory Commission Document Contml Desk. Emergency Action Level Revision, September 24,1992 O
Catvert Cidfs EAL Basis Document N.3 June 15,1993 A
NATURAL PHENOMENA Ernemency Classification Level: ALERT ( App _i_ cable Operational Modes: ALL Calvert Chffs initiatino Condition: NA1 Natural Phenomena Affecdng Safe Shutdown NUMARC Recoonition Cateoory Hazards and Other Conditions Affecting Plant Safety NUMARC Initiatino Condition: HA1 Natural and Destructue Phenomena Affecting the Plant Vtal AnBa Bemer: Not Applicable NUMARC Generic Basis:
< Generic > EAL 1 should be based on (sita-specirc) FSAR design basis. Seismic events of this magnitude can cause damage to safety functions. < Generic EAL 3> should be based on (site-specife) FSAR design basis. Wind loads of this magnitude can cause damage to safety functions. < Generic EAL 2> should specify (site-specifc) stmetums containing systems and funcuons mquimd for safe shutdown of the plant.
~~t Genenc EALs 4, 5, and 6 am addmssed under C 043, Destructive Phenomena Affecting Safe Shutdown.~~
~~< Generic > EAL 7 covem other (SceSpecife) phenomena such as flood.~~
Each of these < generic > EALs is intended to address events that may have msulted in a plant vital area being subjected to forces beyond design hmits, and thus damage may be assumed to have occurned to plant safety systems. The initial
~~report" should not be interpmtad as mandating a lengthy damage assessment befom classifcation. No attempt is made in th<ese Generic > EAL ~~to assess the actual magnitude of the damage.~~~~
Escalation to a higher emergency class, if appmpriate, will be based on < Equipment Failum, Electrical, Fission Pmduct Barrier Degradation, Radioactivty Release, or SEC> Judgement ICs. Bant-Soecific information: Cahert Cliffs EALs am based dimatly on NUMARC and include the independent Spent Fuel Storage installation (ISFSI), as appropriate. EAL 1 is written as: Seismic Event Causing Gmund Acceleration GREATER THAN O.08g Horizontal Or OD53g Vertical l This EAL addresses the Operating Basis Earthquaka (DBE) as desonbed in UFSAR Section 2.6.52. i l l I i Calvert Chffs EAL Basis Document N.4 June 15,1993 l
NATURAL PHENOMENA EAL 2 is wntten as: Venfed Report to Control Room of Visible Damage to Safe Shutdown Equipment \ ! Venfication of damage can be by physical observation, or by indications of degraded equipment performance in the Control Room or at local control stations. EAL 3 uses a sustained wind speed of 90 MPH to addmss high winds striking the Plant Vital Ama as recommended by NUMARC. This speed is chosen to assure that the wind speed is within the design capability of the meteomlogical tower. Thus, EAL 3 is wrcen as: , Sustained Wind Speed GREATER THAN 90 MPH (40 Meters /Second) for AT LEAST 15 Minutes The duration of 15 minutes is selected to indicate sustainedaesids and to preclude wind gusts. Wind speeds are also pmvided here in meters /second for dose assessment input. The convemion equation is as follows: m es " = 58 8P.gnmetem 90 =528 d = 1 hour mile 3600 seconds 32808 feet second Per UFSAR Section 2.8.3.6,the still water level used for intake Structure analysis is 17.6 feet MSL This is above the top of the range of the Tide Level Recorder (CRR-5195). The top of the Traveling Screen cover housings is about 18 feet MSL EAL 4 indicates achieving the design water level. Thus, EAL 4 is wrnen as: Bay Water Level At Or Above the Top of the Traveling Screen Cover Housing Per UFSAR Section 2.8.3.7, the predicted extreme low tide is -3.6 feet MSL and the plant is designed to safety operate at an extreme low water level of -6.0 feet MSL EAL 5 is based on the lower elevation. l Thus. EAL 5 is wrcen as: ! I Bay Water Level is AT LEAST 6 Feet Below Mean Sea Level as indicated by CRR 5195 Tde Level Recorder Or Physical Measurement I I Physical measurements may be mquired if the Tide Level recorder is out of service. For example, the Surveillance Test Procedures provide a way to determine Bay level using a rope. kurce Documents / References / Calculations:
1. Updated Final Safety AnaF/sis Report

2. Operating instruccon (Oil 46, Seismic Measurement Equipment

3. BGSE Drawing 60-220E (M-31). Equipment Location Service Building, Water Treatment Area & Intake Structure Section "JJ j i

4. BG&E Intemal Memorandum, JE Thorp to RE Denton, Emergency Action Level Review Criteria. June 1, 1990 I Calvert Cliffs EAL Basis Document N.5 June 15,1993 l
~~n i~~
~~. -- .- - ..-,e .~~
O 1 l h OTHER HAZARDS 1 O 1 l I O
DTHER HAZARDS Eme_rge_n_cy Classifcation Level: UNUSUAL EVENT Appicable t Doerational Modes: ALL _Cabert Cirffs Initiatino Conddon: OU1 SEC Judgement , NUMARC Recoonition Category Hazards and Other Conditions Affeaing Plant Safety 1 NUMARC_Initiatino Condstion: HU5 Other Conditions Existing Which in the Judgement of the Emergency Diremor Warrant Declaration of an Unusual Event Bamer: Not Applicable NUMARC Generic Basis: This < Generic > Eal is intended to address unanticipated conditions not addressed explicitty elsewhere but that warrant declaration of an emergency because condtions exist which are believed by the <SEC> to fall under the Unusual Event emergency class. From a broad perspective, one area that may warrant GC> judgement is related to likely or saual breakdown of site specific event mitigating actions. Exampled to consider include inadequate emergency response procedures, transient response either unexpected or not understood, failur or unavailabilty of emergency systems during an accident in excess of that assumed in accident analysis, or insufficient availability of equipment and/or y support personnel. Specific examples of actual events that may require <SEC> judgement for Unusual Event declaration are listed here for consideration. However, this list is by no means all inclusive and is not intended to limit the discretion of the site to prwide further examples.
Aircraft crash onsite

Train derailment oraite

Nearsite explosion which may adversely affect normal site actMties.

Nearsite release of toxic or flammagle gas which may adsersely affect normal site activities

Uncontrolled RCS cooldown due to Secondary Depressurizouon it is also intended that the <SEC's> judgement not be limitad by any list of events as defined here or as augmented by the site. This list is provded solely as examples for consideration and it is recognized that adual events may not a} ways follow a preconcebed description.
Santfoecific Information: Site Emergency Coordinator (SEC) is the tde for the emergency director function at Calvert Cliffs. Thus, the EAL-is written as: Any Condition Which in the SEC's Judgement Indicates Potential Degradation in the Level of Safety of the Plant in this manner, the EAL addresses conditions that fall under the Notdcabon of Unusual Event emergency classAcation description contained in NUREGO654, Appendix 1 that is retained under the NUMARC methodology. Calvert Cirffs EAL Basis Document June 15,1993 0:1
DTHER HAZARDS Source Dop_uments/Refemnces/ Calculations:
1. Emergency Response Plan ,

2. NUREGOSS4/EMA-REP-1, Criteria for Pmparation and Evaluation of Radological Emergency Response Rans and Pmpamdness in Support of Nuclear Power Rants. Revision 1. October 1980, Appendix 1 5
~~O 7~~
~~.I i~~
l
\
l 1 Calvert Cirffs EAL Basis Document 02 June 15,1993 l
DTHER HAZARDS Emergency Classification Level: UNUSUAL EVENT , Appligable Operational Modes: ALL Calvert CMfs initiatina Condition: OU2 Toxic or Rammable Gases NUMARC Recognition Cate_ggry Hazan$s and Other Conditions Affecting Plant Safety NUMARC Initiatino Condition: HU3 Release of Toxic or Flammable Gases Deemed Detrimental to Safe Operation of the Plant Darrier- Not Applicable fiUMARC Generic Basis: This IC is based on releases in concentrations within the see boundary that will affect the health of plant personnel or affecting the safe operation of the plant with the plant being within the evacuation area of an offsite event (La, tanker truck accident releasing toxic gases, etc.) The evacuation area is as determined from the DDT Evacuation Tables for Selected Hazardous Materials, in the DOT Emergency Response Guide for Hazardous Materials. < > PlantcSpecific Information: For the purposes of this IC, Halon (such as is discharged by the fire suppmssion system) is not toxic. Fire { s suppressant discharge can be lethal if it reduces oxygen to low concentrations that are immediatay dangerous to life and health \\DLH). Rm suppmssant discharge into an area is not basis for emergency classification under this C unless: [1] Access to the affected area is requimd, and (2) Fim suppmssant concentration results in conditions that make the ama inaccessible (i.e, IDLH). EAL 1 is wrnten as: OnSte Toxic or Flammable Gas Release Which in the Shift Supervisor's Judgement Could Potentially Degrade the Level of Safety of the Plant EAL 2 is written as: Notification of a Nears. te Release That May Require Evacuation of Plant Personnel This EAL addmsses relaases that could originate from the Cove Point Liquid Natural Gas Plant. Source Documents / References / Calculations:
~~1. Abnormal Operating Procedures~~
~~ADP-11, Control Room Evacuation and Safe Shutdown Normre Condtjons~~
~~2. Updated Final Safety Ana9 sis Report O~~
Calvert Cliffs EAL Basis Document D:3 June 15,1993
[TIHER HAZARCE Erne _rgency Classification Level: UNUSUAL EVENT _CN epplicatde_DAerational Mode _s: AU. L Qatyert Cliffs Initiatina Conditi90: OU3 Destructive Phenomena [JUMARC Recoanition Category. Hazards and Other Conditions Affect.ing Plant Safety NUMARC initiating Condition: HU1 Natural and Destructive Phenomena Affecting the Plant Pmtected Area parrier: Not Applicable NUMARC Generic Basis: The pmtected area boundary is typically that part wit.hin the security isolation zone and is defined in the site security plan. Generic EALs 1,2, and 3 are add, ssed by C NU1, Natural Phenomena
< Generic > EAL 4 is intended to address such items as plane or hercopter crash, or on some sites, train crash, or barge crash that may potentia!y damage plant structums containing functions and systems mquired for safe shutdown of the plant. If the crash is confirmed to a'fect a plant vital ama, the event may be escalated to Alert.
For < generic > EAL 5, ony those explosions of suffcient forte to damage permanent structures or equipment wthin the protected area should be considered. As used here, an explosion is a rapid, violent, unconfined combustion, or a catastruphic failure of pressurized equipment, that potentially imparts significant energy to near' by stmetures and materials. No attempt is made in this EAL to assess the actual magnitude of tre damage. The occurrence of the explosion with reports of evidence of damage (ag., deformation, scorching) is sufficient for the declaration. The <SEC> also needs to consider any securcy aspects of the explosion, if applicable.
< Generic > EAL 6 is intended to addmss main turbine rotating component failures of suffcient magnitude to cause observable damage to the turbine casing or to the seats of the generator. Of major concem is the potential for leakage of combustible fuels (lubricating oils) and gases (hydrogen cooling) to the plant environs. Actual firss and flammable gas buildup om appropriately classified via <lCs IU1 and DU2>. This < generic > EAL is consistent with the definition of an Unusual Event while maintaining the anticipatory nature desired and recognizing the risk to non-safety related equipment. Escalation of the emergency classifcation is based on potential damage done by missiles generated by the failum <>. or in conjunction with a steam Denerator tube mpture, for a PWR. These latter events would be classified by the <Radioscuvey Release, Equipment Failure, or Fission Product Barrier Degradation > ICs. ,
l
~~< Generic > EAL 7 covers other (9t.tSpecife) phenomena such as hurricane, flood, or seiche. These Eats can ;~~
also be precursom of more serious events. In particular, sites subject to severs weather as defined in NUMARC l station blackout initiatives, should include an EAL based on natJvation of the severe weather mitigation procedures (eg, precautionary shutdowns, diesel testing, staff calkuts, etc.) < > ; i O Cah/ert Cliffs EAL Basis Document o.4 June 15,1993 i i
OTHER HAZARDS Plant &pecific Information: These EAls em based directly on NUMARC and include the Independent Spent Fuel Storage installation (ISFSI). Releases of flammable gases and fires wthin the protected area. am addressed by other EAL and are thus not I separately addressed due to turbine failum. EAL 1 is wntten as: Nuclear Security Report of an Explosion Wchin the Plant Protected Area Dr Wchin the ISFSI Protected Area i EAL 2 is wntten as: i Visible Damage to Safe Shutdown Equipment Dr to Permanent Equipment or Structums Wehin the ISFSI Pmtected Area ' EAL 3 is wrcten as: Turbine Failure Causing Observable Casing Damage Observable is used to indicate that such damage can be medily seen and does not mquire special equipment or techniques to see or measure. EAL 4 is wntten as: Vessel or Vehicle Collision Causing Observable Damage to Safe Shutdown Equipment EAL 5 is written as: Vessel or Vehicle Collision Causing Observable Damage to Structures Containing Dry Stored Spent Fuel EALs 4 and 5 addmss airplane, helicopter. barge, boat, train, car, or truck collisions into equipment required to achieve or maintain safe shutdown or with the Horizontal Storage Modules and their associated structural supports. These EALs do not include vehicle crashes with each other, damage to office structures, damage to equipment not mquired to achieve or maintain safe shutdown. or damage to structures that are not mquimd to maintain the integnty of the dry spent fuel stored in the ISFSt. Safe Shutdown areas and equipment of concem are identified below. Areas of Concem for Safe Shutdown
Contml Room

Electrical Penetratson Rooms

Control Room HVAC Room

Auxiliary Feedwster Pump Room

Cable Spmading Room

Charging Pump Rooms

Cable Chases

Diesel Generator Rcoms

Swechgear Room o Refueling Water Tank (RWT) 11(21)

ECCS Pump Room

Condensate Stt/49e Tank (CST) 12

Service Water Pump Room

Pretreated Water Storage Tank (PWST) 11(21)

Component Cooling Pump Room

Fuel Dil Storage Tank (FDST) 12

Main Steam Penetration Room This hst of Safe Shutdown areas is displayed on the EAL Tables to assum that all amas related to Safe Shutdown are considered by the SEC.
Cakert Cliffs EAL Basis Document D:5 June 15.1993
DTHER HAZARDS EAL 6 is wntten as: Flooding of Rooms Containing Safe Shutdown Equipment Flooding indicates that the net water flow into the room resuts in elevated water levels, may be more than available dmin capacity, and if continued, can prevent operation of equipmont in the room. Thus, minor water level increases that may resut in wet floors and do not pose a challenge to equipment operation are not included in this EAL Areas containing equipment required for Safe Shutdown are listed above. The rooms located below MSL include the ECCS Pump Rooms and the Charging Pump Rooms. The Shutdown Cooling Heat Exchangers are also located in the ECCS Pump Rooms. Such flooding can resut in a potential degradation in the level of safety of the Calvert Cliffs plant and is therefors included in this EAL Source Documents /Referenceg/ Calculations:
1. Updated Final Safety Analysis Report

2. BG&E Drewing 6CL220E (M31). Equipment Location Service Building, Water Tmatment Area & Intake Structure Section "JJ'

3. BG&E Drawing B3-278f., Plan Auxiliary Building Restricted Access Area EL (-)8'O*,(-110'O' And (-115'O'

4. Letter, G.C. Creel (BGCE) to U.S. Nuclear Regulatory Commission Document Control Desk. Emergency Action Level Revision, September 24,1992 b
t Calvert Cliffs EAL Basis Document 0:e June 15,1993
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~~l oegaav op;s 3yl ges!s oconwau3 OE rnua LS' L66E~~
~~- UTHER HAZARDS s Emergency Classification Level: ALERT Apphcable Operational Modes: ALL Calvert Cirffs initiatino Condition:~~
, OA2 Toxic or Rammable Gases Affecting Safe Shutdown NUMARC Reco_gnition Cateoore Hazards and Other Condrtions Affecting Plant Safety NUMARC Initiatino Condition: HA3 Release of Toxic or Rammable Gases Within a Facilty Struccum Which Jeopardtzes Operation of Systems Required to Maintain Safe Operatbns or to Establish or Maintain Cold Shutdown Barrier: Not Applicable NUMARC Generic Basis: This C is based on gases that have entered a plant structure affecting the safe operation of the plant. This C applies to buildings and areas contiguous to plant Vtal Areas or other significant buildings or amas (ie Service Water Pumphouse). The intant of this C is not to include buildings (le., warehouses) or other amas that are not contiguous or immediateFy adjacent to plant Vtal Areas. It is appropriate that increased monitoring be done to ascertain whether consequential damage has occurred. Femfarinn to a higher emergency class, if appropriata, will be based on < Electrical. Equipment Failure RadioactMey Release Fission Product Barrier Degradation, or SEC Judgement Os> <> ( Plant 40ecific information: For the purposes of this C. Halon (such as is discharged by the fire suppression system) is not toxic. Fire suppressant discharge can be lethat if it mduces oxygen to low concentrations that are immediately dangerous - to hie and health (IDLH). Rre suppressant discharge into an area is not basis for emergency classification under this C unless: (1) Access to the affected area is requirtd, and (2) Rre suppressant concentration results in conditions that make the ama inaccessible (Le., DLH). t Thus, the EAL is written as: Toxic or Rammable Gas Making Safe Shutdown Areas inaccessible This EAL also addresses releases that could originate from the Cove Point Uquid Natural Gas Plant. The areas of concem for safe shutdown are identifed below. & a Calvert Cirffs EAL Basis Document 0.8 June 15,1993
OTIC tt*2ARDS Amas of Con m for Safe Shutdown
Contml Room

Electncal Penetration Rooms

Control Room HVAC Room

Auxiliary Feedwater Pump Room

Cable Spmading Room

Charging Pump Rooms

Cable Chases

Diesel Generator Rooms

Switchgear Room o Refueling Water Tank (RWT) 11(21)

ECCS Pump Room

Condensate Storage Tank (CST) 12

Service Water Pump Room

Pretreated Water Storage Tank (PWST) 11(21)

Component Cooling Pump Room

Fuel Oil Storage Tank (FDST) 12

Main Steam Penetraton Room This list of Safe Shutdown amas is displayed on the EAL Tables to assure that all amas mlated to Safe Shutdown are considemd by the SEC.
~~Source Documents /Refemnces/ Calculations:~~
~~1. Updated Final Safety Analysis Report O~~
O l 1 1 1 l 1 x Calvert Cirffs EAL Basis Document 0:9 June 15,1993 l
DTHER HATARDE l Emergency Classification Level: ALERT O _ Applicable Opgrational Modes: ALL Catvert Clrffs inciating Condcion: DA3 Destructive Phenomena Affecting Safa Shutdown NUMARC Recoonition Catego_ry Hazards and Other Conditions Affecting Rant Safety NUMARC Initiatino Condtion: HA1 Natural and Destructbe Phenomena Affecting the Plant Vtal Ama Sarrier: Not Applicable NUMARC Generic Basis: Generic EALs 1,2, and 3 are addressed under C NA1, Natural Phenomena Affecting Safe Shutdown.
~~<3er9ric> EAL 4 should specify the types of instrumentation or indications including judgement which are to be used to assess occurrence.~~
~~43eneric> EAL 5 is intended to address such items as plane or helicopter crash, or on some sites, train crash, or barge crash into a plant vital ama.~~
< Generic > EAL 6 is intended to address the threat to safetymlated equipment imposed by missiles generated O by main turbine mtating component failures. This (sit & specific) list of areas should include all safety < elated equipment, their contmis, and their power supplies. This EAL is, therefore, consistent with the definition of an ALERT in that if missiles have damaged or penetrated areas containing safety < elated equipment the potential exists for substantial degradation of the level of safety of the plant. < Generic > EAL 7 covem other (Site Specific) phenomena such as flood.
Each of these < generic > EALs is intended to address events that may have resulted in a plant vital area being subjected to forces beyond design limits, and thus damage may be assumed to have occurmd to plant safety systems. The initial " report" should not be interpreted as mandating a lengthy darr, age assessment prior to classification. No attempt is made in th<ess> EAL to assess the actual magnitude of the damage. Ermfebn to a higher emergency class, if appropriate, will be based on < Equipment Failum. Electrical, Fission Pmduct Barrier Degradation, Radioactivity Release, or SEC> Judgement Cs. Plant-Specific information: The Calvert Cliffs EALs are based on report to the contml room of damage affecting safe shutdown functions. EAL 1 addmsses airplane, helicopter, barge, boat, train, car, or tmck colisions. This EAL does not include vehicle crashes with each other, damage to office structures, or damage to structures that are not safety-related. Thus, EAL 1 is wrtten as: Vessel or Vehicle Collision Affecting the Ability to Achieve Or Maintain Safe Shutdown Cakert Chffs EAL Basis Document 0:10 June 15,1993
~~. . MM _ _ _ _~~
EAL 2 is wrcten as: Missiles Affecting the Abilty to Achieve Dr Maintain Safe Shutdown > EAL 3 is wrcen as: Flooding Affecting the Ability to AchWve Or Maintain Safe Shutdown Determination of whether the collision, missiles, or flooding am atfecting abilty to achieve or maintain safe shutdown is determined by physical observation, or by Control Roony' local control station indications. Damage to one train of safe shutdown equipment when other mdundant equipment / trains em operable does not affect the ability to achieve or maintain safe shutdown (e.g., damage to an auxiliary feedwater pump when at least one other auxiliary feedwater pump is operable). The list of amas of concem for Safe Shutdown am shown below and are prominently displayed on the EAL Table. i Areas of Concem for Safe Shutdown
Contru! Room

Electrical Penetrauon Rooms

Contml Room HVAC Room

Auxiliary Feedwater Pump Room ,

Cable Spreading Room

Charging Pump Rooms

Cable Chases

Diesel Generator Rooms

Swtchgear Room

Refueling Water Tank (RWT) 11(21)

ECCS Pump Room

Condensate Storage Tank (CST) 12

Service Water Pump Room

Pmtreated Water Storage Tank (PWST) 11(21)

Component Cooling Pump Room

Fuel Oil Storage Tank (FDST) 12

Main Steam Penetration Room This list of Safe Shutdown amas is displayed on the EAL Tables to assum that all areas related to Safe Shutdown are considemd by the SEC.
~~Source Documents / References / Calculations:~~
~~1. Updated Final Safety Analysis Report O~~
Calvert Cirffs EAL Basis Document 0:11 June 15,1993
I l nieen Ho.ZARDS l 1 Emeroency Classifcation Lgvel: ALERT l Applicable Operational Modes: ALL l Cahvert Cirffs initiatina Conddon: DA4 Control Room Being Evacuated NUMARC Recoanition Cateco_ry Hazards and Other Conddons Affecting Plant Safety NUMARC Initiatina Conddon: HA5 Control Room Evacuation Has Been initiated Barrier: Not Applicable NUMARC Generic Basis: Wrth the control room evacuated, additional support, monitoring and dimetion thmugh the Techn' cal Support Center and/or other Emergency Operations Center is necessary. Inability to establish plant contrul fmm outside the control room will escalate this event to a Site < Emergency >. Plant-Specific Information: This EAL addresses events requiring evacuation of the Control Room such as fire or toxic gas release that make the Contml Room uninhabitable and transferring of control to local stations outside the control room. ADP-9A
~~+ [ fire conditions) and ADP-11 (non4im conditions) control actions for Control Room evacuation and reestablish control of the plant.~~
~~Thus, the EAL is wntten as: Contml Room Evacuation initiated per ADPGA or ADP 11 ; Source Documents / References / Calculations:~~
~~1. Abnorma! Operating Procedums~~
ADP-9A, Control Room Evacuation and Safe Shutdown Due to a Sevem Contml Room Rre ,

ADP-11, Control Room Evacuation and Safe Shutdown Nonfire Conddons i
O Catvert Cliffs EAL Basis Document 0-12 June 15.1993 i s
UTHER HAZADOS Emergen_cy.Qiassification Level: SITE EMERCENCY Applicable Operational Modes: ALL Calyert Cidfs initiating Condition: OS1 SEC Judgement NUMARC Recognition Catego_ry Hazards and Other Conddons Affecting Plant Safety NUMARC Initiatino Condition: HS3 Other Conditions Existing Which in the Judgement of the Emergency Dimator Warrant Declaration of Site
~~<E>mergency Barrier Not Applicable NUMARC Generic Basis:~~
This < Generic > EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because condracns exist which are believed by the <SEC) to fall under the emergency class description for Srte < Emergency >.
+
Plantepecific information: Srta Emergency Coordinator (SEC) is the title for the emergency dimctor function at Calvert Cliffs. r Thus, the EAL is written as: ( Any Condition Which in the SEC's Judgement Indicates Loss or Potential Loss of Two Fission Product Barriers in this manner, the EAL addresses conditions that fall under the Site Emergency classification and is consistent
]
with the Fission Product Barrier Degradation EAL Table. 1 Loss means that a severe challenge to a fission product barrier (Fuel Clad, RCS, or CNTMT) exists such that the , barrier is considemd incapable of performing its safety function. I Potentialloss means that a challenge to a fission product barrier (Fuel Clad, RCS, or CNTMT) exists such that the barrier is considered degraded in its ability to perform its safety function. hu_rce Documents /Refemnces/ Calculations. l
1. Emergency Response Plan

2. NUREGO654/ FEMA-REP-1, Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Pmparedness in Support of Nuclear Power Plants, Revision 1 October 1980, Appendix 1 l
l Calvert Cirffs EAL Basis Document 0:13 June 15,1993 I I
i DTHER HAZARI6 l Erne _mency Classification Level: SITE EMERGENCY p Applicable Operational Modes: ALL Qalvert Cliffs Initiatina Conddon: OS2 Contml Room Has Been Evarw*ad AND Tmely Rant Control Can NOT Be Established NUMARC Recoanition Category Hazards and Other Conditions Affecting Plant Safety RUMARC initiatino Conddon: HS2 Control Room Evacuation Has Been initiated And Plant Contml Can Not Be Established Sanieg Not Applicable , NUMARC Generic Basis: Expedtious transfer of safety systems has not occurred but fission product barrier damage may not yet be indicated. (Site-Specific) time for transfer based on analysis or assessments as to how quicidy control must be reestablished without core uncovering and/or core damage. This time should not exceed 15 minutes. In cold shutdown and refueling modes, operator concem is directed toward maintaining core cooling such as is discussed in Generic Latter 88-17,
Loss of Decay Heat Removal". In power operation, hot standby, and hat shutdown modes, operator concem is primarily dimcted toward maintaining critacal safety functions and thereby assuring ,
fission product barrier integrity. Escalation of this event, if appropriate, would be by < Fission Product Barner j Degradation Radioactivity Release, or SEC> Judgement Cs. o i f 1 Plant Specific Information: ' ( l This EAL addmsses events requiring evacuation of the Control Room such as fire or toxic gas release that make I the Control Room unhabitable and transferring of control to local stations outside the control room. ADP-9A (fire condtions) and ADP-11 (norWire conditions) control actions for Control Room evacuation and re-establish control of the plant. l l An analysis was performed of how quicidy control must be reestablished at Calvert Cliffs without core uncovery l or damage to develop an appropriate site-specific EAL A RETRAN simulation shows that the steam generators i go dry at about 47 minutes for the ADP-9 (station fire) scenario. RCS pressure reaches the lowest pressurizer l l safety valve setpoint soon thereafter, Restoring feedwater within 45 minutes assures that RCS pressum remains below the safety valve setpoint thus avoiding inventory loss. The maximum time allowable to restore RCS l inventory for Appendix R (station fire) scenarios is 90 minutes. Site Emergency declaration at 30 minutes and ; 60 minutes for inability to restore feedwater and RCS make up respectrvely thus consttutes a conservative action for emergency response. Thus, the EAL is wntren as: Control Room Evacuation initiated AND Ether of the Following:
Inabilty to Establish Auxiliary Feedwater to AT LEAST Dne Steam Generator Wthin 30 Mineas

Inability to Establish Reactor Coolant Makeup (Charg:ng Pump Flow) Wahin 60 Minutes l
1 Calvert C!rffs EAL Basis Document 0:14 June 15,1993
J UTHER HAZARDS S2yrce Documents /Refemnces/Calgu!@.cs: f' 1. Abnormal Operating Procedums (
ADP-9A, Control Room Evacuation and Safe Shutdown Due to a Semre Contml Room Fire

AOP-11. Control Room Evacuation and Safe Shutdown Non Fim Condmions
~~2. Letter, LB. Russell (BG&E) to James H. Joyner (US. Nuclear Regulatory Commission Region I), Ememency Action Level Review Meeting. June 6,1991 i~~
em O 4 1 a CaWrt Cirffs EAL Basis Document 0:15 June 15,1993 i
OTHER HAZARDS Emergency Classification Levet GENERAL EMERGENCf A;pJigable_O.perational Modes: ALL Calvert Cliffs Indatino Condition: DG1 SEC Judgement NUMARC Recoonition Catecore Hazards and Other Conddons Affecting Rent Safety NUMARC Initiatino Conddon: HG2 Other Conditions Existing Which in the Judgement of the Emergency Director Warrant Declaration of ) General Emergency Barrier: Not Applicable NUMARC Generic Basis: This < Generic > EAL is intended to address unanticipated conditions not addressed explicitly elsewhere but that warrant declaration of an emergency because conddons exist which are believed by the <SEC> to fall under the General Emergency class. <> Plant-Soecific information: Site Emergency Coordinator (SEC) is the tde for the emergency dima:or function at Cakert Cliffs. Thus, the EAL is written as: 1 f-b Any Condition Which in the SEC's Judgement Indicates Potential for Radiological Releases Requiring Off- j Site Protectrve Actions I in this manner, the EAL addmsses conditions that fall under the General Emergency classification description l contained in NUREGO654, Appendix 1. Source DocumentsfReferences/ Calculations:
1. Emergency Response Plan

2. NUREGO654/ FEMA-REP-1. Criteria for Pmparation and Evaluation of Radiolog' cal Emergency Response Rans and Prepamdness in Support of Nuclear Power Rants Revision 1, October 1980, Appendix 1 l
~~CaNert Cliffs EAL Basis Document 0:16 June 15,1993 1~~
~~-i}}~~

ML20046A478

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