ML20086J842
| ML20086J842 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 11/10/1983 |
| From: | Counts M SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | |
| Shared Package | |
| ML20086J835 | List: |
| References | |
| PROC-831110-01, NUDOCS 8401250318 | |
| Download: ML20086J842 (70) | |
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"Y NUCLEAR OPERATIONS PROCEDURES EMERGENCY PLAN PROCEDURES TABLE OF CONTENTS ;
DATE PROCEDURE NUMBER PROCEDURE TITLE REVISION ISSUED EPP-001 Activation and Implementation of anergency Plan 6 11/9/83 EPP-002 ' Communication and Notification 8 11/9/83 EPP-003 In-Plant Radiological Surveying 3 6/1/83 EPP-004 Out-of-Plant Radiological Surveying 4 5/25/83 . EPP-005 ,0ffsite Dose Calculations 5 11/9/83 EPP-007 Environmental Monitoring 1 7/27/82 EPP-009 On-Site Medical 2 7/27/82 EPP-010 Personnel / Vehicle Decontamination 3 11/9/83 EPP-011 Personnel Search and Rescue 2 7/27/82 EPP-012 Onsite Personnel Accountability & Evacuation 4 5/25/83 EPP-013 Fire Emergency 1 5/25/83 EPP-014 Toxic Release 1 7/27/82 EPP-015 Natural Emergency (Earthquake, Tornado) 3 11/9/83 - EPP-016 Emergency Facilities Activation & Evacuation 4 11/9/83 EPP-017 Post-Recovery and Re-Entry 4 11/9/83 EPP-018 Emergency Training and Drills 3 11/10/83 EPP-019 Bnergency Equipment Checklist 4 6/1/83 EPP-020 Emergency Personnel Exposure Control 2 5/25/83
- Safety Related Procedure' -
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- p. 1 l PROCEDURE DATE NUMBER PROCEDURE TITLE REVISION ISSUED EPP-021 Activation of the Early 'darnirig Siren *
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.l EPP-022 Verification of Communications Operability 2 5/25/S3 l
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, b[]/ 11/3/83 LIST OF EFFECTIVE PAGES r
PAGE REVISION
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'I 10 5 11 5 . {' , ATTACHMENTS Attachment I Pages 1 thru 13 6 D Attachment II 5 i Q,i 1 Attachment III 5 4 0 b 4 e X
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ENERGENCY ACI!ON LEVELS SUP9tARY (:m . EPP-001, ATTACHMENT I I i i~ , PAGE 1 of 13, REVISION 6 , q, , , SYSTEM SITE- GENERAL , , i: i ifdVSUE EVENT AtERI D40RGENCY fMFRGENCY {
- ,.l INVOL VID '
REACIOR INITIATIE CumITION INITIATING CO@lTION INITIATING COW IIION INITIAi!1G CONDITION COOLANT EECEEDING TECHNICAL SPECIFICATION RAPID FAILURE OF SEVERAL STEAM KNON1 LOSS-OF-COOLANT ACCIDENT SMALL APD LARGE LOCAs W!iH FAILtRE (- l I SYSTEM PRIr4ARY TO SEC0 W ARY OR PRIMARY GENERATOR TUBES (e.g., SEVERAL (LOCA) CREATER THAN CHARGING PUW OF ECCS 10 PERFORM LEADING 10 [' j i SYSTEM LEAK RAIES HLMDRED GPM PRIMARY-10-SECONDARY CAPACITY. SEVERE CORE DEGRADATION OR ICLT. i
! LEAK NATC). ULTIMATE FAILURE OF REACTOR s l Detection Method: Detection Methods _ BUILDING POSSIBLE FOR IELTDONN j.
Primary to secondary leak rete Detection hthod: Pressurizer low pressure reactor SEQUENCES. { , greater than 1 rye total for more trips and f'
+ than 4 hours or greater than 500 Pressurizer low pressure alarm anc Pressurizer low pressure safety Detection Nthod: [
y d per steam generator as reactor trips argi injection signal, and_ identified by daily RCS leakage Pressurizer low le. vel alarms M High Reactor Building pressure, Safety injection signal with
, evaluations o_r_ RM-A9, Gl9A, C198, or Gl9C valid and reactor trips M ,
Primary system leak rate greater high alarms M - High Reactor Building susp level 1) Statue lamps indicate safety { . injection and RHR pumps not than those specified in Technical Steam generator water level and Specification 3.4.6.2 as identifiet rapidly increasing in one or enore Higte Reactor Building humiditys runnings or
- i, by daily RCS leakage evaluation. steam generator (s), fslling in the and 2) Flow indicators for Safety .
others; a!nff RM-A2 high alare Injection System read zero; asg
- 1) > 0 pressure boundary leakage. RM-L3, or RM-LIO valid high alara RHG-5, RHG-7, RHG-18, high alare; >
and RM-A2 high alare; i I 2) > I gpa unidentifled for more and ------------------ [ than 4 hours Possible lifting of steam INITIATING CONDITION l- l 4' 3) > 10 gpa identified RCS leakage generator PRV's and/or safety RAPID FAILURE OF SEVERAL STEAM J
' ! for more than 4 hours. valvee. GENERATOR TUBES (SEVERAL HUWRED -------------,----
- 4) > 33 ye controlled leakage (at GPM PRIMARY-10-SECtyOARY LEAK INITIATING CONDIi!ON 2235 f,20 peig for more than 4 ----------------- RATE) WITH LDSS OF 0FFSITE POWER. SMALL LOCA AND INilIALLY SilCCESSFiR.
] ECCS. SUBSEQt[NI FAILtRE OF } , hrs.) INITIATING CONDITION
- i 1 s 5) > 1 r,pe RCS pressure isolation REACTOR COOLANT LEAKAGE RATE Detection Wthod REACTOR DUILDING HLAT REMOVAL valve per table 3.4-1 for , EXCEEDS 50 GPM. SYSTEMS OVER SEVERAL HOURS COULD gre_st_er t_h_en_A_ hours. _ _ _ _ _ Pressurizer low pressure alare and LEAD TO CORE MELT AND POSSIBLE q
INITIATING CONDITION Detection Methods reactor trip, g pressurizer level FAILtRE OF THE REACIOR 10!LDING.
- , A! NORMAL REACIOR COOLANT TEW . rapidly decreasing and ,
1, j AND/0R PRESSURE OR A0 NORMAL FUEL 1) Excessive Reactor Mdce-m to Pressurizer low-level alareg and Detectinn hthnds ; IEMPERATURE THAT WOULD INDICATE A the Volume Control Tank RM-A9, Gl9A, Gl98, or Gl9C valid Pressurizer low prerssure reactor i
- 2) Operational Leakage Test high alarms and_ loss of 115KV and trip; and ;
j } LDSS OF SUBC00 LING MARGIN OR OVERPRESSURIZAi!ON. Procedure STP-ll4.002/ 230 KV ESF potential lights, eM Pressurizer low pressure sarety , l- Detection Wthod Steam Generator water level injection signal d i j,! 1) Reactor Coolant Tavg greater rapidly increasing in one or more 1) RHR flow indicators show zero than 590*Fg or, _ steam generators, falling in the flow after shift to RIR is ,
- 2) ACS pressure greater than 2310 others: M attempted and for greater than 2 !
;. i psigs g RM-L3, or RM-LIO valid high alarms hr. eubsequently.nnd RCS , l
- 3) RCS pressure less than 1765 temperature rising, d f a_rd
- 2) Reactor Building spray and psig; and core exit temperature ,
3 greater than 620*F ma indicated by Possible lifting of steam genera- Reactor Building air handling l a vahd incore thermocoupleg 3 tor PRV's and/or safety valves. system fall to function. ; i 4) Inadequate Core Cooling Alarm. * + I. '. 4 i I i r
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[HERGENCY ACTION LEVELS SumARY EPP-001, ATTACllHENT I , I
- PAGE 5 of 13, REVISION 6 l
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SITE GENERAL i j~ SYSIEH tlERT FHfRGENCY FHERGENCY [ l INVOIVfD INISUAL EVENT INiilATING CONDITION INIi!ATING CONDIIIGN j f fift INITIA7ING CONDITION INii!ATING CONDIi!ON POSSIBLE FUEL DAMAGE. DEGRADED CORE Willi POSSIBLE LOSS LOSS OF IWO OF IllREE~f!SSION { l DAMAGE FUEL DAMAGE INDICATION. Of C00 LADLE CE0HEIRY. PRODUCT DARRIER W!iH POTENTIAL { l ' Detection Hethnd LOSS Cr IHE THIRD OARRIER (e.g., l Detection Method .. ! Detection Methods LOSS Of FUEL INIEGRIIY Ato PRIMARY COOL ANI 000NDARY AND HICil I
,I RM-L1 High altra; and Laboratory RM-L1 High Alarm; M Laburatory analysis which indicates dose m alysis which indicates dose POIENilAL FOR RADI0 ACTIVITY I
equivalent I-131 activity equivalent I-131 activity Wide Range ihot above 700*f, d RELEASE FROH CGNIAINtENI l concentration 1 30 Ci/ge. in the concentration 1 300 C1/gm in the Thot and icold
; primary coolant, primary coolant. rapidly diverging ( i rapidly Detection Methemt: i i
increasing) or no T acrras core;
' and RM-L1 offseale (greater than 1) LOCA and Containment Isolation i 10G cpm) statu U ights indicate incomplete isolation:
- INITIATING CONDII M -or-HAJOR DAMAGE TO MORE THAN ONE SPENT FUEL ASSENDLY IN REACIOR 2) LOCAs M
' ...._ ..__ ...... DUILDING OR FUEL HANDLING DUILDINT RM-A2 high alarms and INii!ATING CONDITION LEADING 10 CLAD RUPIURE (e.g., IU4G-7 and/or f04G 18 reading
- greater than 10 4 R/hr; g FUEL HANDLING ACCIDENT W!ill LARGE 003ECT DAMAGES FUEL OR WATEl<
RELEASE Of RADIDACTIVITY TO LOSS DELOW TUEL LEVEL). Reactor Building pressure . REACIOR OR FUEL HANDLING DUILDING. greater than 30 psig fur at Detection Method: least 2 min.
- Detection Methods or Observations of major damage to !
- 3) Or'each of containment Oteervation of damage to spent more than one spent fuel assembly integrity (Reactor Building) ,
fuel assembly; and or and RM-L1 off-ocale (> 10 6 Spent fuel pool water below fuel epm)
- 1) RM-A2 high alarm and/or level d RHG-17A,178 high alarm after accident in Reactor Building; 1) RM-A2 and/or RHG-17A,178 higt or alarm for accident in Reactor
- 2) RM-A6 high alarm; or Building orr RHG-8 high alare; 2) RM-A6 and/or RHC-8 high clarm after accident in fuel for accideat in fuel Handling Handling Duilding. Building I \
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&P-001, ATTACitqENT I, EMERGENCY ACTION LEVELS StMtAP'i ,
s PAE 6 of 13, REVISION 6
,*j. ,,
SITE
. GENERAL .
SYSTEM {,
',j INVOLVED IMUSHE EVENT AlERI EKHGENCY
, INiilATING CONDITION -
EMIRGENCY l STATION INITIATING ComITION INITIATING CONDITION . INT 11ATING ED W IIl0N ( LOSS Of OffSITE POWER AND LUSS Of trl2 OT OffSITE POWER AND LDSS OF TAILtRE Of OFFSIIE AND ONSIIE
.i POWER total LOSS OF OffSITE POWER OR LDS$
ONSiW A-C POWER FOR MORE THAN 15 POWER ALONG WIIH 10lAL LDSS OF Of GNSITE A-C POWER CAPA8ILITY ALL ONSITE A-C POWER FOR MDRE ' THAN 5 MINUTES HIN. . EMERGENCY FEEDWAIER MAKEUP. Detection Method CAPABILITY FOR SEVERAL HOURS. Detection Method COULD LEAD TO EVENTUAL CDRE MELI , _ Detection Methods. ,
- 1) Loss or 115 KV and 230 KV ESF Afe POSSIBLE FAILiftE OF IHE ,
f'. '
- 1) Loss o* 115 KV and 230 KV ESF Potential Lights
- 1) Loss of 115 KV and 230 KV ESr. Potential Lights and
. Potential Lights REACIOR Bult01NG. .
and and 2) A or B Trn. Blackout Seq. Detection Mettal
- f 2) A or B Trn. Blackout Seq. Ir.t. Alarm (
l', 2) A or B Trn Blackout Seq. Int. Int. Alarm ' and 1)- Loss or 115 KV and 230 KV ESF. f Alarm and 3) Both Diesel Generators Potential Lights. . Inoperable. M i j - ', or 3) Both Diesel Generator j Inoperable ----;------------- 2) A or 8 Trn. Blackout Seq. Int.
.----------------- INITIATING CO W ITION Alarm j
- 3) Both Diesel Generators !
l ;d Inoperable for) I hour ' INITIATING CONDITION LOSS Of ALL VITAL ONSITE D-C an_d LOSS Of ALL DNSITE D-C POWER FOR POWER FOR MDRE THAN 15 MIN. 3) Both Diesel Generators ' A PERIOD GREATER THAN 5 MINUTES. Detection Methn<f r Innperable.
- I I i g.
' Detection Method: 1) D-C bus undervoltage alarms or 4) Steam driven Emergency , ],' all buses! Feedwater pump rails to atert j. g and is inoperable for several ; j' hours. t 1) D-C bus undervoltage alarme 2) A80V ESF Channel A or 8 Lose # on all buses: M of DC Alarm
- f. ,
- 2) 480 V EST Channel A cr 8 Loon 3) DG A or B Loos of DC Alarm l of DC Alarm d . '
i INITIATING COWITION
- 3) DG A or D Loss of DC Alarm LDSS OF RNCTIONS NEEDED FOR i
PLANT HOT SHUIDOWN. l*: Detection Hethoit:
.' ) Inability to establish .
charging pump injections g . a' + t inability to establish !* emergency Feedwater Flow. 2 I or
- 2) liit:1 System not functional l I f F
e i
- 2. '.. .
e L m i p - 1 D4ERGENCY ACTION LEVELS SufHARY EPP-001, AllACllMENT I .. f' i PAGC / of 13 REVISION 6 SITE GENEPJL SYSTEM # AtERI EHFRGENCY fHfPGry*Y INVOL VE D INISUAL EVENT INITIATING CIM)til0N INI T I Al lNf'.- 00NDIi10N INIlIATING CON 0! TION INITIAi!NG CONDIIiON ENGINELRLD SAFEGUARDS. ((CS INITIATED (C0!NCIDENT Willi ALL ANNUN(lIAIOR ALARMS LOST. ALL ANNUNCIA10RS AND PLANT IP.ANSILNI-INITIATLD DY LOSS Of []
" , COMPUTER LOST FOR HORE THAN 15 rEEDWATER AND CONDENSAIE SYSilHS
, . POSITIVE flND!NG IlMI INITIATION 15 NOT SPURIOUS). , Detectinn Nethods HINuiES AND PLANT TRANSIENT (fli!NCIPM. lll AT P[HDYAL SYSif H)
INITIATED OR IN Pit 0GRESS. - FOLLOWED BY_FAlltME Of DIERGINCY ,l
> Detection Het y Observation by operator / TEEDWAIL!t SYSIEH TOR EXIENDED Det ec t ion Het had: PER100. CDRE KLiING POSSIBt E IN Valid Safety injection Signal SEVERAL HOURS. tlLIIMAIE FAILIAtE ,
t .
~
.- - --- Observation of event. Of REACIOR IMJILDING POSSihlE If ,
' ~
I' IN!!!ATING CONDITION CORE WLIS.
~[ - -- .C ' --
EVACUATION of CONTROL ROOH ---------------- INITIATING CONDITION . ANTICIPATED OR REQUIRED WITH INITIATING C0t0ITION Detectinn Hetint . FAILtEE Of A PRESSURIZER OR STEAM CONTROL Of SHUIDOWN SYSTEH EVACUAIION Of CONTROL ROOH A W ESTABLISilED FRtN LOCAL SIAi!O!.'S. CONTROL Of SituTDOWN SYSTEMS'NOT Reastor trip on steam flow >
- 0.ENERATOR SAFE 1Y OR Rd IEF. VALVE II '
RESCAT (EXCEEDING N0HMAL WEEPAGt'). ESTABLISHED FROH L0c't STAI!ONS teedwater flows M IN 15 HIN. Ducreasing wide-range steam Detection Method: Detection Hettwwf generator levels toward off-scale r low on all steam generators! Pressurizer c,r7 steam generator Detection Hethods and
- safety or relief valve opens and 1) EntGgency feedwater flow *
+ then falls to reset as indicated indi'catora indicate zero flow
- by: '% Same as initiating event. 2 min. efter required or [) Velid'open indication of Same as initiating condition. 2) Status lamps indicate -; '
' Pressurizer relief or aarety emergency feedwater painpa not :
- valvo position indication running 2 min. after require >dl (i lights 'md I.
or 3) Emergency feedwater cannot be E valid acoustical monitor restored within~Y) min. Indication. or
- 2) Visual and/or atalible '
indication at vent stacks of open steam generator aafety or , relief valves of Excess feedw?ter flow to and ii steam flow from affected
. generator. .
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.! .. EPP-001, ATTACIMENT I EMERGENCY ACTION LEVELS SutNARY ' . 4
=[ PAGE 8 of 13, REVISION 6 i
'l ) '
- j-
. SITE. GENERAL
- 4 I i. SYSTEM .
ALERT. 1HERGLNCY EME NCfNCY
. INVOLVfD LBRISUM. EVENT -
. INITIATING COWITION .
ENGINEERED INITIATING CO W ITION INITIATING CONDITION TRANSIENT REQUIRING OPERATION Of
. LOSS of ENGINEERED SAFELY FEATURE REACTOlt COOLANT PU W LOCKED ROT 00 SAFEQUARDS SHUTDOWN SYSTEltS WITil FAILURE TO '!
ItMCTION REQUIRING TECHNICAL - , WITH FUEL DAMAGE.
. (CON'T) TRIP WHICH RESULTS IN CORE DAMAGE }-
! SPECIf1 CAT 10N SHUTOOWN. .(TECHNI- Olt ADDITIONAL FAIListE Of CORE - ! 4 CAL SPECIFICATIONS 3.3.2) . Detection Hethod: (
Reactor coolant pusy auto trip COOLING AND MAKEUP SYSTfMS WHICil 4,i COULD LEAD TO C0ftE HELT. (' alarm; and 3' Detection Methods '
}
- 1) Engineered Safety features Reactor trip on low i i coola.it . flows- g Detection Methods Activation System found inop-
* .erable. Reactor coolant pump phase over
---- current relay actuation: M Reactor remains critical after
! ---- --.---- attempted trip g . I INITIATING C0 m lTION valid Iti-1 Alarm. l INDICAll0NS OR ALARHS ON PROCESS 1) RHL-1 alarm, o_r_
3 OR EffltfMT PARANEIERS NOT FUNCTIONAL IN THE CONTROL R00H TO -- - - ---------
- 2) Flow indicators on safety.
[!- INITIATING CONDITION
, AN EXTENT REQUIRING SilOT00WN AS PER TECHNICAL SPEClflCATIONS: OR, LOSS Of ALL FUNCTIONS NEEDED FOR injection system and RHR systeme [
PLANT COLD SHUIDOWN. show zero flow with saft-'w ( 3' 0THER SIGNIFICANT LOSS Of injection initiated. or ! ASSESSENT OR CONNtNICATION (? CAPABILITY. Detection Nethods i
- 3) Status lights show safety !
RHR system not function 1 and injection system and RHR pumps t
- Detection Method 1) Loss of not running with safety ~ {
4 RMietion Monitoring Systems E 2: -inability to reject heat to the Injection initiated. l-
+ Loss of all onsite meteorological ' condenser. i' s
l information g imavailability of ; offsite meteorological
, l - information. -
or 6
- 3) Significant loss of
- ' communication capability offsite. ----------------- -[ l' INITIATING CO W ITION FAILUNE Of THE R E OR PROTECT!06 I SYSTEH TO INITIATE AND C0HPLETE A h TRIP WHICH DRINGS THE REACTOR ,' SUBCRITICAL. l . Detection Methods - r
; t.
. Reactor remains critical after 1
trip initiation.
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. D4ERGENCY ACTION LEVELS StM4ARY EPP-001, AITACt94ENT I &.
,4 PAGEL9 af 13, REVISION 6 ;
,j . .
e. SYSTEM
~~
SliE. GENERAL-INilSUN. EVENT 4 ERT D9ERCrNCY D#hr.ENCY f
, INVOL VED .
FIRE INI11ATING C0tDITION INITIkTING,f9!DIT ION INITIATING CONDlIION_ .
~
!~...
. FIRE LASTING NORE THA,. 15 HINS. FIRE POIENTIALLY AffECTING flRE. Aff ECTING SAFETY TRAINS OR - - sf *
~
SAFETY SYSIEMS. FUNCTIONS.
Detection Method: - Detection Methods _' ;..' Detection tiethod: i Olmervation or fire detection - .' '
; . device eterm with confirming . 0bservation of fire that.could . 0bservation of major fire that .
; ,e' observation indicating a fire affect ona or more ' safety - defeats both trains of a safety systems, ayetem of function. ;
lasting more than 15 min.' i
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'I1 ENERGENCi ACTION LEVELS SUtttARY &P-001, ATIACllHENT I .. f PAGE 10 of 13, REVISION 6' j SYSTEM SITE GENERAL f
- INJSUAL EVENT M ERT DtERCENCY ENERCfNCY_
INVOLVED t SECURITY INIIIATING CDNDITION INITIATING CONDITION INITIATING CONDITION INITIATING CONDITION" -{ h SECURITY TilREAT OR AliEMPTED ENTRi DNCOING SEVERE SECURITY TilREAT SECURITY TilREAT INYDLVING SECURITY TilREAT REsut. TING IN LOSS OR ATTEMPTED SA00iACE. IHMINENT LOSS Of PilYSICAL CONTROL Of PilYSICAL CONTROL Of IllC , Detection Hethod: Of THE PLANT. FACILITY. i i Detection Method: Security safeguards contingency ; event which results in Detection Hethod Detection Mettwf: l
- 1) Report to the Control Room by adversaries commandeering an i observer, a.eu of the plant, but not Physical attack on the Plant Physical attack un the Plant has ,!
i j.
;~ . or impacting shutdown capability. Involving imminent occupancy of resulted in occupation of the
- 2) Alarm of Intbgrated fire and control rone and auxiliary control room said the auxiliary i.
Security Panel. shutdown panels. shutdown panets by unauthorized j. personnel. > k
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y . m/ ,t n EMEh6ENCY ACTION LEVELS SutHARY EPP 001,.AITACINENT I . ..
' PAGE 11 of 13,~ REVISION 6 ,,
SITE GENERA *. SYSTEM . . FHERCfNCY
. INVol.VED INJSUM EVfNT 'AtERT EMERGENCY ,r NATURAL' . INiilATING CD W ITION I
INITIATING CONDITION' INITIATING COMitTION ..
- I'
, PilEN0HENON
? .NAIURAL EVENTS SEVERE NATURAL EVENT NEAR SITE. SEVERC NATURAL PilEN0E NON BEING-EXPERIENCED OR PROJECTED WITH
- a. EARTHQUME PLANT NOT IN COLD SirJTDOWN.
l.
- b. TORNADO ON SITE a. EARTHQUAKE CREATER TilAN THE
- c. HURRICANE NEAR SITE- (00E) LEVEL .
- a. EARTHQt1*E CREAIER THAN SSE
' LEVELS.
- b. TORNADO STRIKING FACILITY 8
- c. SYSIAINED HilRRICANE WINOS b. SYSTAINED WINDS IN EXCESS Of Detection Methods GREATER THAN 75 Wil 100 Wil DN SITE. ,
' Detection Methods
- a. Salam b Recording System Detection Hethods ,
s' Start indication. a). Seismic Responso Spectrum *
- b. Otservation or event. Trouble Annunciator a. . Reactor Building foundation
- c. Observation of event. b) Otmervation or event Seismic Switch (SSE)
.j, c) Heteorological Manitoring .
1 System Data or Weathe.- Burent b. Meteorological.Honitoring
! Report. System Data.
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- ENERGENCY ACTION LEVELS SUM 4ARY EPP-001, ATIACilHENI I t
. PAGE 12 of 13, REVISION 6' j j" GENLRAL. 1 4 SYSIC4 SITE AtERT -THERGENCY 'INERGENCY !.
!' INVOLVfD LMUSUAL EVFNT INITIATING EONDITION f HANHADE INITIAing CDNDITION INITIATING COMilTION
. Pl'ENOMENON .
.OTHER ltA2ARDS BEING EXPERIENCED ,,
i GTHER HAZARDS BEING EXPERIENCED 011 OTHLR HAZARDS DEING EXPERIENCED OR PH0JECIED WITH PLANT NOT IN OR PROJECTED WlICH llAVE A COLD SitUTDOWN: ! L' PROJECTED MtICH llAVE THE PoiENTIAI . FOR ENDANGERING THE FACILITY SIGNIFICANT POTENTIAL FDR a. AIRCRAFT CRASH INTO VITAL l -
~- AFFECilNG PLANT SAFETY. STRUCTURES.
- a. AIRCRAFT CRASH ON FACILITY. b. HISSILE OR EXPLOSION IMPACT b I a. DNSIIC AIRCRAFT CRASH.
- b. HISSILE ,IFPACTS ON FACILITY ON FACILITY RENDERING SEVERE I
' /, b. DNSITE TRAIN DERAILHENT. **' WITH RESULTANT MAJOR DAMAGE. DAMAGE TO SHUTDOWN EQUIPMENT f
- c. . DNSITE EXPLOSION (EXCLtDING PLANNED AcilVITIES). ' c. . KNOWN EXPLOSION AT FACILITY J c. ENTRY Of IDXIC OR FLAHMABLE j
- d. NEAR OR ONSITE 10XIC OR -RESULTING IN MAJOR DAMAGE TO GASES INTO / ITAL AREAS WlICH j '
+ FLAMBLE GAS RELEASE Of A PLANT SIRUCIURES OR INVOLVE A SIGNIFICANI j HAGNiiUDE THAT IHREATENS EQUIPMENT. OEGRADATION Of PLANT PERSONNEL. d. ENTRY INTO FACILITY ENVIRONS SAFETY.
! e. TUROINE GENERATOR ROTATING GT 10XIC OR FLAT #4AOLE GASES Detection Method:
; COMPONENT FAILURE CAUSING IN CONCENTRATION Wi!CH EXCEEI:
RAPID PLANT Situ!DOWN. THE LIMIIS Of FLAHMAGILITY Oli a) Aircraft crash causing
}
{ IOXICITY. , damage or fire ins !
; Detection Hettwxt: e. TUR0!NE GENERATOR FAILURE 1) Reactor Building; nr i
} , .
CAUSING CASING PENE1 RATION. 2) . Control Rooms or
- a. 0bservation of event. 3) Auxiliary Buildings or l
{, b. Observation of event. Detectinn Method: 4) fuel llandling Buildinggor '
- c. Observation of explosion or 5) . DG Buildings E warning from of faite. a) Observation of aircraft crest 6) Intermodlate Building ,
l* d.. Observation of release or into Plant structures. 7) SW Intake Structures I warning from offsite. b) Observation of missile .
; e. Turbine trip and observation impacts on Plant structures b) Loss of functions needed for
- of turbine malfunction or or components. hot shutdown. (see page 6 of '
> Galure. c) Observation of damage by this Attachment) explosion. . .
) .
c) Entry of toxic or flammable d) Observation or warning from gases into: outside the Plents detection of gases (using portable 1) Control rooms g instrumentation) which exist 2) Cable sprea. ding roomst oj in concentrations which 3) Reactor Buildings 02-exceed the limits of 4)' Switchgear rooms or 6 flammability or toxicity. 5) Safe chutdown panelag g e) Turbine trip and observation 6) Emergency diesel of penetration of casing, generator roomas as detected by portable in- ': i strumentation and which ren-dors a train of a safety - l l related system inoperable. ; i-
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WERGENCY ACTION LEVELS SutNARY EPP-001, ATTACl#4ENT I -' 'ly fe. PACE 13 or 13 REVISION 6 .k. l SITE Gt'NERAL - f-l SYSTEM 40 AL ERI ENERGENCY DtERGENCY INVOL VED LNISuAL EVENT ODfER INITIATING CONDITION INITIATING CONDIi!ON INITIATING CDNDITION LOSS Of CONTAINtENT INTEGRITY OillER PLANT COPDI'i!0NS EXIST TilA1 OTHER PLANT CONDITICNS EXIST THA1 REQUIRING TECIINICAL SPECIFICATION WARRANT Aci!VAi!0N Of IECitNICAL WARRANT ACTIVATION OF EMERGENCY , ; SituiDOWN (IECitNICAL SPECIFICATION Sul' PORT CLNTER AND PLACING FACILITIES AND RADIATION 3.6.1.1) EMERGENCY Ol'ERATIONS FACILITY MONITORING TEAHS AND A f PERSONNEL ON STAND BY PHECAUTIONARY PUBLIC WARNING. Detect ton Hethod:
- 1) Any automatic containment Detection Method: Detection Hethmis f isolation valve found to be t j inoperable g As determined by Emergency As determined by Emergency j 2) Reactor Dullding air lock Director. Director.
inoperable E , i
- 3) Penetration (s) fall leak test 2 g
(as specified in Technical J Specifications) :
; INIIIATING ND!IION : ! I' TRANSPORTATION Of DVEREXPOSED l r
AND/0R CONIAMINATED, INJURED
'f INDIVIDUAL FROM SliE 10 it0 SPITAL.
Detection .4ethewl Sume as initiating condition. j INITIATING CONDIi!ON GillIR PLANT CONDITIONS EXIST .i. REQUIRING OilfER illAN NORMAL PLANT ; SitulDOWN AND REQUIRING INCREASED li AWARENESS ON IllE PART Of STATE 4 Off!CIALS. ' i~ Detection Method: As determined by tourgency I-Director. , s
. .e.
__ - , . . . . _ _ _ _ . ___...__._._m . . _ . . ___;____...____-.._.._. . _ _ _ _ _ .
=
j
- j. SOUTH CAROLINA ELECTRIC AND GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION q . .
() NUCLEAR OPERATIONS NUCLEAR OPERATIONS COPY No. . . . . . . g?5. 7.............
- EMERGENCY PLAN PROCEDURE EPP-002 COMMUNICATION AND NOTIFICATIONS
. REVISION 8 NOVEMBER 3, 1983
. SAFETY RELATED
- r. )
(A
- v. /
u __
#!? f3 DISCIPLINE SUPERVISOR ' Date 11/ W rn APPROVAL AUTHORITY ' datei' NOV 9 O DATE ISSUED RECORD OF CHANGES CHANGE TYPE DATE DATE CHANGE TYPE DATE DATE APPROVED CANCELLED NO. CHANGE APPROVED CANCELLED NO. CHANGE ,
T 1 f' .% ig-fk
~
-~ - ._-. _ _ ,
.s
- ~ ~ _ _ _ _ _ _ . .
I EPF -602 PAGE i REVIS' ION 8
^ . .
11/3/83 1 LIST OF EFFECTIVE PAGES PAGES -REVISION i 8 11 7 111 7 1 7 2 7 8
. 3 7 4 7 5 7 ATTACHMENTS Attachment I, Pages 1 & 2 7 c-I Attachment.II, Pages 1 &-2 7 Attachment IIIA 8 Attachment IIIB 8 Attachment IIIC 8 Attachment IIID, Pages 1 & 2 8
-Attachment IVA, Pages 1 - 4 7 Attachment IVB, Pages 1 - 3 7 Attachment V, Pages 1 - 3 7
-Attachment VIA 7 Attachment VIB 7 Attachment VIC 7 Attachment VII, Pages 1 & 2 7 Attachment VIII, Pages 1 - 4 8 1
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EPP-002 ATIACI-MDTP IIIA PAGE'1 OF 1 p '
. ~.
REVISION 8
^
INITIAL IDTIFICATICN UNUSUAL EVDTP PJGER PERSON . AGENCY / PERSONNEL M EBER FDMB SITE N0rrmtv TIME
- 1. Management Duty Supervisors (per Duty Roster)'-
Ollie Bradham 603 798-1359- 4040
- John Connelly 613 798-2719 4257 t
Ken Woodward 663 798-7761 4220 Bruce Williarrs 623 781-7214 4221 , 4140
'Mel Browne 633 732-0897 Mike Quinton 643 8-9-356-1980 4044 Bob Croley 653 781-7069 4042
~ .-
'(. \ 2. *S.C. Dept. of Health & 758-5531
(, Environmental' Control, Bureau or of Radiological Health 758-5548 N/A
- 3. Brerg. Preparedness Division 758-2826 N/A
- 4. *NPC' See Attachnent IVA of EPP-002 for NRC Initial Notification
- 5. Media Coord. (Mgr., Nuclear Info.)
Hczne Office
- 1) Becky McSm in 703 8-9-359-3510 5049
- 2) 1st Alt. - U. M. Lide, Jr. 772-1830 749-3420 771-6804 6.-Site Services Manager-Bob Stough 781-8711 345-5422 r 781-9128 7.-Inst'.tute of Nuclear Power Operations (INPO)
(404)953-0904 N/A (404)953-3600 .y
- 8. American Nuclear Insurers (ANI) See Attachment IVB of EPP-002 for ANI Notification
}OTIFICATIONS CDMPLEIE:
- Dedicated Line Also Exists. 'O ONE a- _
_4 - g
. _ _ . _ . _ . _ . _ . __ . _ ... . .. .- ..._____ .._-,__ _ ; _ _ . __....L.__
EPP-002 INITIAL IDTIFICATICN ATTACSENT IIIB PAGE,1 OF 1 ALERT REVISION 8 V PMER PEltC.ON MENC?/PERSOthrI, NUMBER EDME SITE BDTIFIID TILE
- 1. Managenent Duty Supervisors (per Duty Roster) .
'Ollie Bradham 603 798-1359 4040 John Connelly 613 798-2719 4257 Ken Woodward 663 798-7761. 4220 Eruce Willia:rs 623 781-7214 4221 ~.
i Mel Browne 633 732-0897 4141 Mike Quinton 643 8-9-356-1980 4044 Bob Croley 653 781-7069 4042 tDTE: To contact additional Emergency Response Personnel refer to Policy Menorandum No. 25.
- 2. *S.C. Dept. of Health & 758-5531
^ Enviroturental Control, Bureau or-I. ), of Radiological Health 758-5548 N/A
- 3. Energ. Preparedness Division 758-2826 N/A
- 4. *NRC See Attachment IVA of EPP-002 for NRC Initial Notification
- 5. Media Coord. (Mgr., Nuclear Info.)
HcIne Office
- 1) Becky McSwain 703 8-9-359-3510 5049
- 2) 1st Alt. - W. M. Lide, Jr. 772-1830 748-3420 771-6004
- 6. Site Services Manager-Bob Stough 781-8711 345-5422 781-9128
- ;. 7. Institute of Nuclear
, Power Operations-(IN m ) (404)953-0904 N/A (404)953-3600
- 8. Fairfield Punped Storage Facility 345-5830 N/A .
- 9. American Nuclear Insurers (ANI) See Attachnent IVB of EPP-002 for ANI Notification N7fIFICATIONS CEMPLEII:
DATE
- Dedicated Line Also Exists. CALLER'S SIGNA?JRE
**9l. W.
- _ ..- * .- -m4. p. ., A. _ ~ , , . . ,,em.. , , ,p,
, - ee e 6
- EPP-002 ATTACHMENT IIIC INITIAL _ NOTIFICATION PAGE 1 OF 1 REVISION 8
* " SITE EMERGENCY" m, / - NAME OF PAGER PERSON AGENCY / PERSONNEL NUMBER HOME SITE NOTIFIED TIME
- 1. Management Duty Sugarvisors (per Duty Roster) .
~
^ Ollie Bradham 603 798-1359 4040 John Connelly 613 798-2719 4257 Ken Woodward 663 798-7761 4220 4 Bruce Willliams 623 781-7214 4221 Mel Browne 633 732-0C97 4141 Mike Quinton 643 8-9-356-1980 4044 Bob Croley 653 781-7069 4042 NOTE: To contact additional Emergency Response Personnel, refer to Policy Memorandum No. 25.
- 2. *S.C. Dept. of Health & .
r's Environmental 2ontrol, 758-5531
'\ _,1 Bureau of Radiological or Health 758-5548 N/A
- 3. Emerg. Preparedness Division 758-2826 N/A
- 4. *NRC See Attachment IVA of EPP-002 for NRC Initial Notification Requirements
- 5. Media Coord. (Mgr.,
Nuclear Info.) Home Office 1)- Becky McSwain 703 8-9-359-3510 5049
- 2) ist Alt. - W.M. Lide, Jr. 772-1830 748-3420 771-6804
- 6. Site Services Manager -
Bob Stough 781-8711 345-5422 781-9128
- 7. Institute of Nuclear Power Operations (INPO) (404)953-0904 N/A (404)953-3600
- 8. Fairfield Pumped Storage ~
- 3. Facility 345-5830 N/A
- 9. American Nuclear Insurers See Attachment IVB of EPP-002 for ANI Notification.
s_/ NOTIFICATIONS COMPLETE: Caller's Signature Date
- Dedicated Line Also Exists.
a
. . . - ~ . . . ~ . . - . - - _ _..~.-.-a-... . . . _ - _ . . - _ . . . _ . ~ . . , ~ . . . - _ . - _ . -
EPP-002 ATTACHMENT IIID PAGE 1 OF 2 INITIAL NOTIFICATION REVISION 8 (T
' ._ ,i
" GENERAL EMERGENCY" NAME OF
" PAGER PERSON
; AGENCY / PERSONNEL NUMBER HOME SITE NOTIFIED TIME
- 1. Management Duty Supervisors (per Duty Roster)
. 1) Ollie Bradham 603 ,
798-1359 4040 , 2 .' John Connelly 613 798-2719 4257
- 3) Ken Woodward 663 798-7761. 4220
}. 4) Bruce Williams 623 781-7214 4221
- 5) Mel Browne 633 732-0897 4141
- 6) Mike Quinton 643 8-9-356-1980 4044 i
- 7) Bob Croley- 653 781-7069 4042 _
NOTE: To contact additional Emergency Response Personnel, refer . to Policy Memorandum No. 25.
- 2. '*S.C. Dept. of Health &
Environmental Control, 758-5531 Bureau of Radiological or Health 758-$548 N/A
- 3. *NRC' See Attachment IVA of EPP-002 for NRC Initial Notification Requirements
- 4. Media Coord. (Mgr., Nuclear Info.) Home Office
. 1) Becky McSwain 703 8-9-359-3510 5049
- 2) ist Alt. - W.M. Lide, Jr. 772-1830 748-3420 771-6804
- 5. Site Services Manager -
Bob .7 tough 781-8711 345-5422 781-9128
, 6. Fai rf. eld Pumped Storage Facility 345-5830 N/A
- 7.
- Emergency Preparedness Division-Adjutant General's Office 758-2825 N/A
~ (State EOC) i'
7
- Dedicated Line Also Exists.
w e- w -- y- 9 ,g a's -ye> e- p =w v & - m .
,. _ . . . - _ . . ~ . . . . ~- . - - -- .. . - . . - - . . - . . - - - . . . _ . . . - - - . .
EPP-002 ATTACHMENT IIID PAGE 2 OF 2 ['N - - REVISION 8 INITIAL NOTIFICATION
" GENERAL EMERGENCY" NAME OF PERSON -
~
' AGENCY / PERSONNEL NORMAL SITE- NOTIFIED TIME
- 8. *Fairfield County 5-635-5511 N/A 9..*Newberry County 66-276-4295 N/A
- 10. *Richland County 9-254-9296 N/A l 11. *Lexington County 8-9-359-8141 N/A
- 12. Institute of Nuclear Power Operations (INPO) (404)953-0904 N/A
, (404)953-3600
- 13. American Nuclear Insurers See Attachment IVB of EPP-002 for ANI Notification.
^
NOTIFICATIONS COM'PLETE: Caller's Signature Date 1 I
- Dedicated Line Also Exists.
I k 1
. , - . ~- .. . , , . . . . ,
EPP-002 ATTACHMENT VIII PAGE 1 0F 4 REVISION g Emergency Operaticr.s Facility Layout X _.
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1 - COMMAND CENTER (Room 144) 9 - NRC AREA (Room 122) 2 - TECHNICAL SUPPORT AREA (Room 145) 10 - CONFERENCE ROOM (Room 3 - OFFSITE RADIOLOCICAL MONI'*1 RING AREA 143 (Room 145) 11 - COMMUNICATIONS AREA 4 - GENERAL SERVICES AREA (Room 138) (Room 140) - 5 - CONSTRUCTION / REPAIR AREA (Room 138) 12 - REPRODUCTION AND 6 - MEDIA AREA (Room 141) RECORDS STORAGE AREA "I - LOCAL GOVERNMENT AREA 13 - STROAGE AREA i (Room 142) 8 - STATE GOVERNMENT AREA , (Room 124)
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1 EPP-002 . ATTACHMENT VIII
'5J - ' PAGE 2 OF 4
.s REVISION 8 OFF-SITE EMERGENCY TELEPHONE LIST Emergency Operations Facility i
- 1. COMMAND CENTER .
~
5077 . EOF-Media Coordinator (R. McSwain)
;- 5078 EOF-Emergency Control Officer (0.W. Dixon) 5079 . EOF-Offsite Radiological Monitoring i Coordinator (B. Baehr) 5080 EOF-Security Coordinator (J. Harrison)
I 5081 EOF-General Services Coordinator (J. Bailey) 5082 EOF-Construction / Repair Coordinator (J. Woods) 5083 EOF-Technical Support Coordinator (K. Nettles) 5053 URC Representative
. 'i
- 2. TECHNICAL SUPPORT AREA 5075 EOF-Technical Support Engineers 5076 EOF-Technical Support Engineers /Telecopier
- 3. OFFSITE RADIOLOGICAL MONITORING AREA 5072 EOF-Offsite Rad. Monitoring 5073 EOF-Offsite Rad. Monitoring 5074 EOF-Offiste Rad. Monitoring
- 4. GENERAL SERVICES AREA 5067 EOF-General Services Personnel
- L-
- 1 5068 EOF-General Services Personnel .
5069 EOF-General Services Telecopier s
~~ -
EPP-002 - ATTACHMENT VIII PAGE 3 of 4
'" REVISION 3 OFF-SITE EMERGENCY TELEPHONE LIST Emergency Operations Facility
- 5. ' CONSTRUCTION / REPAIR AREA .'
5070 EOF-Construction Personnel
~
5071- EOF-Construction Personnel u. i 6. MEDIA AREA a
- j 50G1 EOF-Company Media Personnel 5062 EOF-Company Media Personnel 5063 EOF-Company Media personnel -
- 7. LOCAL GOVERNMENT AREA
-5064- EOF-Local Government Officials EOF-Local ~ Government Officials
- 5065 EOF-Local Government Officials 5066
- 8. STATE' GOVERNMENT AREA
.5058 EOF-State Government Officials I- 505h EOF-State Government Officials
- 9. NRC AREA
- i 5053 EF-NRC Officials ti 5054 EOF-NRC Officials 5055 EOF-NRC Officials 5056 EOF-NRC Officials q .
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l l
' EPP-0 02 -
ATTACHMENT VIII f7~,
' ,1 PAGE 4 OF 4
-REVISION 8 OFF-SITE EMERGENCY TELEPHONE LIST.
i:
- Emergency Operations Facility 3 '4'
- 10. COMMUNICATIONS AREA l
; . 5051 EOF-Emergency Communicator j .- 11. SECURITY CONTROL POINTS 1
j 5085 DOF-Hallway Area 4 EOF-North Entrance 5086 1 j. 1 1 1 L 9 (
; .J v-O h
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--._-.:..,- . . . ~ . -~ . - - .. . , . . . - . . . - . . . ~ . , . . . . -
SOUTH CAROLINA ELECTRIC AND GAS CCMPANY i VIRGIL C. SU:GIER NUCLEAR STATION O) * *
's d NUCLEAR OPERATIONS
- n NUCLEAn OPERAT copyno..........t*s7..%% *** \ , EMERGENCY PLAN PROCEDURE EPP-005 1 OFFSITE DOSE CALCULATIONS REVISION 5 NOVDIBER 3, 1983 5 SAFETY RELATED O v1A J DISC (PLINE SUPERVISOR as
' Date MA tw //h/Q APPROVAL AUTHORITY Date NOV 9 1993 DATE ISSUED RECORD ':F CHANGES , CHANGE TYPE DATE DATE CHANGE TYPE DATE DATE I NO. CHANGE APPROVED CANCELLED NO. CHANGE APPROVED CANCELLED .
l t I
_ _ _ _ . ._ . _ . - _ . . _ . _ . . . . z .
. , . _ . . _ . , . .._ _ ___,. . _z.._____ _, _ .. _ _ _ , , , _
l EP.'-0 0 5 PAGE i RINISION 5 11/3/83 s.- LIST OF EFFECTIVE PAGES Page Revision L. 5 , 5~
-11 lii- 5 iv Deleted 5
1 5 2
. 3 5 4 5 .
i 5 3 5. 5 E 6 7 5-t ' 8 5
' ATTACHMENTS 4
I II, Pages 1.and 2 4
.III, Pages 1 and 2 4 4
IV i 4
. V 4
VI 4 VII 4 VIII 4 IX-4 X
-4
'XI
' 4
-- XII:
XIII, Pages 1 through 3 4 4 XIV 4 XV 4 XVI 5
.' XVII
'XVIII, Pages 1 through 4 4 XIX, Pages l'through 7 5
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EPP-005 PAGE ii RtVISION 5 r~g , , 11/3/83 i TABLE OF CONTENTS
'# Pace
-List-of Effective Pages i
-1.0 PURPOSE 1 2.0_ REFERENCES 1 .
i 3.0 CONDITIONS AND PREREQUISITES 2 4.0 PROCEDURE 2
. 5.0 BASES 6 ! ATTACHMENTS ATTACHMENT I - Sample form: Meteorology and Source < Term Report Form ATTACHMENT II - Sample Form: Emergency Off-Site Dose Calculation Using Stack Monitor Source Term ATTACHMENT III- Sample Form: Off-Site Dose Calculation Using High Range Monitors 3
ATTACHMENT IV - Sample' Form: Activity Release Status Q) ATTACEMENT V - Sample Form: Out of Plant Monitoring-Dispatch & Data Form ATTACHMENT VI - Sample Form: Dose Off-Site (Doce Rate) Calculation from Environmental Monitoring ATTACHMENT VII -Dose-Concentration Conversion ATTACHMENT VIII-Delta T Atmospheric-Stability Classification ATTACHMENT'IX - Sigma Theta Atmospheric Stability Classification.
-ATTACHMENT X - Meterological Sector Definition ATTACHMENT XI - Correction Factor for High Range Monitors ATTACHMENT XII -Protective Action Guides (PAGS) e at
... A!
- e
, = , g, -e p
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T.~ 2. n.- . _ . ..,..a-.__- _.. ._ _ _ - - .. ; . : . - - . .. _.. _ u . - -a
'. I EPP-005 PAGE iii ,
REVISION 5 g s,1 . , 11/3/83 i TABLE OF COtiTENTS (cont) List of Effective Pages ATTCHMENT XIII- Pasquill-Turner Method for Determining Atmospheric Stability.
-ATTACHMENT XIV- Volume-Pressure Steam Curve ATTACHMENT XV - Containment Leakage Source Term-ATTACHMENT XVI- Calibration Constant (pre HPP-906 implementation) - ATTACHMENT XVII - Calibration Constant (post HPP-906 implementation).
ATTACHMENT XVIII - Projected Whole Body and Thyroid Exposure Rate at 1 Mile ATTACHMENT XIX - Estimation-of Release Rates From FSAR Listed Accidents I
?
ti t' I fi ~i 4 f
!f l3
!f e % y
- k. a 4
EPP-005 REVISION 5 11/3/83
- f 1.0 PURPOSE v
1.1 To provide methods for estimating wholebody and child thyroid dose rates due to abnormal releases of radioactive materials to the environment. 1.2 DTo provide a method of comparing the integrated wholebody , and child thyroid doses to the EPA Protective Action - Guides (PAG's).
2.0 REFERENCES
2.1 " Virgil C. Summer Nuclear Station Radiation Emergency Plan". 2.2 American Society of Mechanical Engineers, " Recommended Guide for the Prediction of the Dispersion of Airborne Effluents", May 1968. , 2.3 Dames and Moore, " Data Reduction Procedure for Virgil C. Summer Nuclear Station", September 1974. 2.4 Slade, D. H., (ed.), Meteorology and Atomic Energy - 1968", TID-24190, Division of Technical Information, USAEC p (1968). t
- ~" Turner, D. B., "A Diffusion Model for an Urban Area",
2.5 Journal of Aeolied Meteorolocy, February 1964. 2.6 U. S. Nuclear Regulatory Commission, Proposed Revision 1 to Regulatory Guide 1.23, " Meteorological Programs in Support of Nuclear Power Plants", September 1980. 2.7 NUREG-0654/ FEMA-REP-1, " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants", December 1980. 2.8 Emergency Assessment and Response Systems, (EARS 1,000 and EARS 9845). Technical Manual. Applied Physics Technology, Inc. 2.9 EPA-520/1-75-001 (Revised June 1980) " Manual of Protective Action Guides and protective actions for Nuclear Plants." 2.10 EPP-002 Communication and Notification. A% m PAGE 1 OF 8 GENERAL REVISION e .n. w w +*
. . _ . . . . _ _ . - . _ ~ . ,_ _ _.
EPP-005
! REVISION 5 11/3/83 -
[~') 2.Il Virgil C. Summer Nuclear Station Final Safety Analysis
%' Report, . Chap'.er 15, Sections 15.3 and 15'.4.
2.12 HPP-906 Preplanned Alternate Method for Monitoring Post Accident Parameters as required by Technical Specifications. 3.0 CON 0ITIONS AND PREREQUISITES 3.1 None s 4.0 PROCEDURE 4 NOTE: This procedure delineates two (2) methods by which offsite doses may be calculated (i.e., computer calculation and manual calculation). Either method, or a combination of both methods, may be used since comparable results can be obtained. 4.1 Overlay Selection 4.1.1 If the computer is not available proceed to step 4.2 to initiate manual calculations. 4.1.2 Use the HP9845C in the Technical Support Center to (~h s- run the menu driven real-time dose projection system. The program is started by following guidance derived through Reference 2.8. 4.1.3 Complete step 4.3 so that the manual method map is available for use in case of computer failure. The map is also useful'in dispatching out of plant monitoring teams. Proceed to step 4.5. 4.2 Manual Method 4.2.1 Stability Class Determination If the meteorological tower is inoperable proceed to step 4.2.2 otherwise, complete Part 1 of Attachment I to obtain meteorology and source term data from control room instrumentation. Method A. (below) is preferred. i 3
'~
PAGE 2 OF 8 GENERAL REVISION
-+.-#,,
EPP-005
'~ REVISION 5 11/3/83 -
'n *
- All meteorological data must be obtained
) NOTE:
' - ' from strip charts by averaging the most previous 15 minutes of data for a parameter.
A. Differential Temperature (a T) 61-10 meters and 40-10 meters Method. *
- 1. Obtain a T information from the Control Room instrumentation.
NOTE: Priority should be given to the 61-10 meter information with the back up being the 40-10 meters.
- 2. If the Control Room instrumentation is
' inoperable, personnel with a portable radio may be dispatched to the meteorological tower to obtain AT information directly from the met tower instrumentation. Keys to towers are available in the Emergency Planning Key Box located in the Emergency Operations Area of the T.S.C.
- 3. Determine the Pasquill Category by using Attachment VIII and select appropriate X/Q overlay found near the plotting map.
(
- 4. Proceed to Step.4.3 for determination of plume location and X/O values.
B. Sigma Theta Method
- 1. Review the wind direction chart located in Control Room and determine wind range for a period of 15 minutes to 1 hour.
'2. Wind range is determined by the difference between the highest and lowest values measured. If the wind direction chart shows one or two spikes, these should be ignored in determining the wind range. If more than two spikes, these must be considered in determining wind range.
- 3. If the wind range is larger than 180
+
dsgrees, use 180* as the wind range value.
+
Q: PAGE 3 OF 8 GENERAL REVISION
& .. .,. w. - . .
" " - " #+ ' +- - 3 A 4 5
' " '
- b
EPP-005 REVISION 5' 11/3/83
- 4. Divide the wind range by six (6).
- 5. Determine the Pasquill Category by using Attachment IX and select appropriate overlay.
4.2.2 Backup meteorological data can be obtained from
- the National Weather Service (NWS).
A. If meteorological tower data is not available call the NWS (see ref. 2.10 for phone number) and obtain windspeed, wind direction, % cloud cover and ceiling height. Also, obtain current forecast and note any fronts that may cause a rapid shift in wind speed or direction. I Complete Attachment XIII " Turner's algorithm" to obtain stability class. 4.3 Location of the Effluent Plume , l 4.3.1 After determination of appropriate overlay in Section 4.2 place overlay to the center of the base map. 4.3.2 Orient the centerline of the overlay along the .
- < s, direction the wind is blowing.
(
'~' Wind directions are given as directions from which NOTE:
the wind is blowing; affected directions are 180* from the wind direction. When communicating wind direction, specify the direction from which the wind is blowing (e.g. wind is blowing from 270') 4.3.3 The conversion from wind direction measured in degrees to wind direction in sectors is presented in Attachment X. 4.4 Manual Dose Calculation , 4.4.1 Complete Part 2 of Attachment I. A. To perform a quick estimate of the projected offsite radiation exposure, complete Attachment XVIII as per the instructions on the back of the form. PAGE 4 OF 8 GENERAL REVISION - l
- . .. _ . . .. .._ _,_._~ -- , . _ - _ ..._ _.. _ .. ._ m.__ -._._ _ _ __
EPP-005 REVISION 5 11/3/83 , [~h *
- 4.4.2 ' Complete Attachment II if radiation monitor RM-A3 i' (Gas) or RM-A4 (Gas) is on scale Compare results .
with the PAGS-(Attachment XII). 4.4.3 If RM-A3 (Gas) or RM-A4 (Gas).i.s offscale or if j there is a steam release complete Attachment III. Compare results with the PAGS (Attachment XII).
- NOTE: Attachments VII, XI and XIV, will be necessary for completing ~ Attachment III.
4.4.4 Complete Attachment XV if either radiation monitor RM-G7 or RM-Gl8'are on scale. Compare results with PAG's (Attachment XII). 4.4.5 If the radiation monitors on the effluent-path are y offscale or inoperable and HPP-906 has not been implemented,Lrefer to Attachment XIX for~ estimated Release Rates and Iodine to Noble Gas Ratios derived , from Reference 2.11. I a) Choose the listed accident and accident assumptions from Att. XIX that most closely matches the event at hand. b) Enter.on line C of Att. II the Noble Gas Release l Rate from the chosen accident in Att. XIX. c) Enter on line I of Att. II the Iodine to Noble Gas Ratio'from the. chosen accident in Att. XIX.. NOTE: For'the Control Rod Ejection Accident and the Loss of Coolant Accident, the Noble Gas Release Rate must be adjusted if containment leakage is other than the assumptions. stated. For the Fuel Handling Accident, the Noble Gas Release Rate must be adjusted if the number of fuel rods damaged is more than the assumption in the realistic case as stated. d) Using Att. II, compute off-site dose projections. i,- Compare with PAG's (Att. XII).. n !!- 4.4.6 Complete Attachment IV to maintain curie accountability. Total activity of iodine, noble gas, particulates, and tritium must be kept current. (Hourly updates if possible) ,
.4 P.
PAGE 5 OF 8 GENERAL REVISION !i ,
~
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l o l EPP-005 l REVISION 5 l 11/3/83 . ; s
- 4.4.-7 Transfer all applicable results to EPP-002 Attachment II for transmittal and documentation purposes.
4.5 Environmental Monitorina 4.5.1 As soon as practical, dispatch monitoring teams to the affected meteorological sectors. Attachment ~ V, should be used to record location and results. NOTE: Dose projections based on source terms should be verified es soon as possible by supportive measurements. Based on these measurements, a reassessment of the PAG's should be taken. 4.5.2 Attachment VI can be used to project doses based on environmental samples. These can then be compared with source term projections. I 5.0 BASES 5.1 This section is included to provide a brief explanation of the sources of constants and graphs used in the procedures and associated attachments. .
,m Until a full class A model is developed,
( ',i A. Step-4.3.1 subsequent to the completion by Dames and Moore of a detailed study of the Summer Station Meteorology, the (X/Q) overlays will be based on a straight line gaussion plume model as described in Reg. Guide 1.145. B. Attachment II, Section I - The Iodine to noble gas activity ratio of 0.3, until measured, is a conservative estimate from Ref. 2.9 page D-59. TMI-2 would suggest that the actual ratio could be 10-5 or-less. A fast and accurate measurement of the actual ratio is important to avoid unnecessary. evacuation orders based on the Iodine PAGS. The calibration constants 3.3x10 -10 and 3.8 x 10 -10 are based on the manufacturer's calibration and various unit conversions as follows: l PAGE 6 OF 8 GENERAL REVISION
&- e eewe-, _ . , ..,
a .* w= m wm _
,-c
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s y* %
, EPP-005
\', REVISION 5 11/3/83 '
=
) '
3.-3X10-10 =1- min x 106 ml x 10-6 Ci 1 u Ci/ml M. sec M y Ci 50X100 CPM 3'.8X10-10 J= 1 min'x 10 6, Fml x 10-6 pl i'44x100 CPM Ci X 1 p Ci/ml
'66_ sec
, *q . . . N.
C.- Attachment'III, Section I - Same comment as B.- . The calibration constants 4.0 x 10-5, 9.5 x ~ , _ 10-5 and 8.5 x $0-4 are based on the i manufacturer's calibration and various conversion as
. follows :'
I
-4.0LX 10-5 = 0'.028 m3 x 106 ml X 10-6 Ci x. 1 min.
Yt3 p pli 66 see 11.7'm 9.5 .x 10-5 2'O.028 m3 x 106 ml x 10-6 Ci x 1 min pl37 _60
. Yt3 mJ sec 4.89 m
%Y,. , .
8.5 x'10-4/= ' 0.028ftT m3 x lo6ml x 10-6 Ci x 1 hr mT ' Ci 3600 E c 9.13x10 u D. Attachment VII - The curves follow the methodology and assumption of Ref. 2.9. The noble gas curve comes
'frem.EQ.'2.9,'the Iodine curve from EQ. 2.13', T
,s* s
~;j x E. Attachment XIII-This description of the Pasquill-Turner method was supplied;by Dames and Moore. The calculation of..the solar altitude was obtained fromt the Alabama Power Company Farley Station EDCM.
.It was verified by comparision with selected values from'the Smithsonian 1leteorological Tables.
'F.- Attachment XII - Protective actions are taken from
- The particulate concentration, 7.5 x 10-7
~
Ref. 2.9. Ci/ml, would lead to an exposure of 2000 MPC Hours in 8 hours. That is, it-is analogcus to a whole body
..L exposure of: REM. Other concentrations are related by.
ratio to- the specified :whole body exposure found in Ref
'2.9.
G. Attachment XIV is a plot of standard-steam table data for saturated steam. T H. Attachment XI is a plot of the ratio of DOSE to
; concentration for mixed ~ fission noble gas.to that g for 133 Xe as a function of time after trip.
1 d'
+
- i. L M m
PAGE 7 OF 8 GENERAL REVISION
*- * ., .-~ ,
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- . . . . . . ~ . . . , . - . . . .-. _ _ . . . _ . _ - _ _,
EPP-005 REVISION.5 11/3/83 I. Attachment XV - the conservative leak rate of 2.56 ft.3/ min is based on the Tech. Spec. allowable limit (Section 3.6.1.2 of 0.2% by weight of containment air per 24 hours at 47.1 ps which is the pressure expected with a steam line break. (1.84E6 ft. 3)x(0.002) x (1 hr.) = 2.56 ft.3 min. / -
~
(24 hrs.) (60 min.) J. Attachment XVI is a plot of calibration constant versus time post-accident for RM-G7 and RM-G18. Data is based on a repo by GAI on RM-G7 and RM-G18 calibration. The calibration constants were derived as follows: ! (1 Ci/ml) x (l_hr.) x (lE-6 Ci) x (lE6ml) x (0.028_m3 ) ~ (Cr mr/hr) (3600 sec) (pCi) (m3) (Ft. J)
=7.78E-5 = calibration constant ,
C' where CF = conversion factor derived from data supplied by GAI. ~ K. Attachment XVII is a plot of calibration constant versus time post-accident for a RM-16/RD-17A monitoring system outside t Reactor Building air lock when HPP-906 has been implemented for either RM-G7 or RM-G18. Data and derivations are as in above. W h J
- s PAGE 8 OF 8 GENERAL REVISION t- . . _ . _
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- EPP-005 ATTACHMENT XIX PAGE 1 OF 7
, REVISION 5 ss ESTIMATION OF RELEASE RATES FROM FSAR LISTED ACCIDENTS STEAM LINE BREAK ACCIDENT Accident Duration: 8 hours Accident Assumptions:
Realistic Case: 0.12% Fuel Dafects prior to accident; 42 GPM Steam Generator Blowdown. Conservative Case: 1.0% Fuol Defects prior to &ecident; 15 GPM Steam Generator Blowdown; 5% Fuel Failure during accident. Noble Gas Release Iodine-to-Noble Rate (Ci/S) Gas Ratio Realistic Case 0-2 hours 3.15E-6 0.288 ' ' 2-8 hours 3.15E-6 0.063 Conservative Case 0-2 hours 2.0218 0.0812 2-8 hours 2.0218 0.0109 CVCS' LETDOWN LINE RUPTURE (~5 w Accident Duration: Up to 1 hour Accident Assumptions: Realistic Case: Shutdown long period of time (up to 9^ days); 0.12%
, Fuel Defects ct time of accident.
Conservative case: Shutdown short period of time (can be operational); 1.0% Fuel Defect at time of accident. Noble Gas Release Iodine-to-Noble l __ . _. Rate (Ci/S) Gas Ratio Realistic Case 0-10 min. . 0278 0.0176 10-20 min. . 0276 0.0174 20-30 min. . 0275 0.0172 30-40 min. . 0274 0.0170 40-50 min. . 0273 0.0168 50-60 min. . 0272 0.0167 y
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- EPP-005 ATTACHMENT XIX PAGE 2 OF 7 m , ,
REVISION 5 i CVCS LETDOWN LINE RUPTURE (Cont'd) Noble Gas Release Iodine-to-Noble Rate (C1/S) Gas Ratio _ Conservative Case 0-10 min. 1.529 .0032 10-20 min. 1.525 .0031 20-30 min. 1.523 .0031 30-40 min. 1.521' .0030 40-50 min. 1.519 .0030 50-60 min. 1.517 .0029 CONTROL ROD EJECTION ACCIDENT Accident Duration: 24 hours I Accident Assumptions: Realistic Case: 0.12% Fuel Defects prior to accident; 100 lbs/ day Primary-to-Secondary Leakage; Dump Steam to Condenser Conservative Case: 1.0% Fuel Defects prior to accident; 1.0 GPM Primary-to-Secondary' Leakage; 10% Fuel Failure . during accident; Dump Steam to Atmosphere. k' Ultra-Conservative Case: 1.0% Fuel Defects prior to accident 1.0 GPM Primary-to-Secondary Leakage; 10% Fuel Failure and 0.25% Fuel Melt during Accident; Dump
-Steam to Atmosphere.
Noble Gas Release Iodine-to-Noble Rate (C1/S) Gas Ratio Realistic Case 0-2' hours 2.57E-5 0.18 2-8 hours 2.46E-5 0.336 8-24 hours 2.37E-5 0.126 Conservative Case 0-2 hours 0.1366 1.15 2-8 hours 0.0916 1.102 8-24 hours 0.063 0.973
. Ultra-conservative 0-2 hours 0.1707 1.031 2-8 hours 0.113 1.0 8-24 hours 0.078 0.876
' ' NOTE:
The above Noble Gas Release Rate is for 0.2% containment leak - rate per 24 hours-(2.5 cfm).
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-e EPP-005 ATTACHMENT XIX PAGE 3 OF 7 e~T , . REVISION E LOCKED PUMP ROTOR ACCIDENT Accident Duration: 8 hours Accident Assumptions: -
Realistic Case: 0.12% Fuel Defects prior to accident. Conservative Case: 1.0% Fuel Defects prior to accident; 10% Fuel Defects during accident. Noble Gas Release Iodine-to-Noble Rate (01/S) Gas Ratio Realistic Case 0-2 hours 3.15E-6 6.54E-3 2-8 hours 3.15E-6 3.62E-3 Conservative Case 0-2 hours 2.'044 0.0114 2-8 hours 2.044 0.0069 . f 'A
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EPP-005 ATTACHMENT XIX PAGE 4 OF 7
. . REVISION 5 LOSS OF COLLANT ACCIDENT (LOCA)
Accident Duration: 1000 hours > Accident-Assumptions: Realistic Case: 100% of RCS Noble Gases and 50% of RCS Iodines .
~
, released. Conservative. Case:. 100% of Gap and RCS Noble Gases and 50% of Gap and RCS Iodines released. Regulatory Guide 1.4 Case: 100% of Core, Gap and RCS Noble Gases and 25% of Core Iodines + 50% of Gap and RCS Iodines released. o Noble Gas
. Release 0-2 2-8 8-24 24-96 4-30 Rate (Ci/S) hours -hours hours hours days .
Realistic' Case 0.00153 0.0015 0.00143 0.0013 0.00088 Conservative Case 0.0467 0.0754 0.0416 0.0339 0.0200
' Reg. Guide 1.4 Case 13.456 7.388 5.032 3.693 2.327 -
(..
- 1
. Iodine-to-Noble 0-2 2-8 8-24 24-96 4-30 Gas Ratio hours hours hours hours days Realistic Case 0.015 0.0131 0.0101 0.0.068 0.0028 Conservative Case 0.047 0.25 0.36 0.304 0.187 Reg, Guide-l.4 Case 0.29 0.35 0.29 0.163 0.085 NOTE: The above Noble Gas Release Rate is for 0.2% containment leak rate per 24 hours (2.5 cfm).
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. . .- - . . . ~ . ~ . . . . - . - . . . . . ., - _ . . . - . . _ . . . _ . _ _ . . . , .
EPP-005 ATTACHMENT XIX PAGE 5 OF 7
^g . . REVISION 5 WASTE GAS DECAY TANK RUPTURE Accident Duration: 0.5 to 10 hours Accident Assumptions: 1.0% Fuel Defects while collecting Waste Gas in tank; accident immediately after isolation from -
~
Waste Gas System; complete contents of WGDT , dumped. Time to Empty Tank Noble Gas Release Iodine-to-Noble (Release Period) Rate (Ci/S) Gas Ratio 0.5 hour 12.23 1.29E-5 2 hours 3.04 1.26E-5
, 4 hours 1.51 1.2E-5 6 hours 0.997 1.*6E-5 8 hours 0.74 1.13E-5 10 hours 0.86 1.lE-5 ,
STEAM GENERATOR TUBE RUPTURE Accident Duration: 8 hours r-s l Accident Assumptions: ( ,. Realistic Caso: 100% lbs/ day (0.0083GPM) Primary-to-Secondary Leakage; 0.12% Fuel Defects; 42 GPM Steam Generator Blowdown. Conservative Cases: 0.1 to 1.0 GPM Primary-to-Secondary Leakage; 1% Fuel Defects; 15 GPM Steam Generator Dlowdown. Noble Gas Iodine-to-Noble Gas Ratio Release 0-2 hours fate2-8(Ci/S) hours 0-2 hours 2-8 hours Realistic Cese 4.45E-2 2.11E-6 4.24E-4 5.98E-3 Conservative Case 2.73 1.42E-4 6.7E-5 3.2E-2 (0.0lGPM) Conservative Case 2.73 1.42E-4 8.35E-5 3.2E-2 (0.lGPM) Conservative Case 2.73 1.42E-4 2.37E-4 3.2E-2
. (1.0GPM) 1 e
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1 1 EPP-005 ATTACHMENT XIX PAGE 6 OF 7 5
'
- REVISION 5 FUEL HANDLING ACCIDENT Accident Duration: 2 hours Accider.t Assumptions:
Realistic Case: Average - Power Assembly; > 100 hours after fuel -
~
assembly subjected to Reactor flux; and one row of fuel pins damaged. Conservative Case: Highest - Pcwer Assembly; 100 hours after plant shutdown, all fuel pins damaged; good reuention of Iodine in Spent Fuel Pool Water. Regulatory Guide 1.25 Case: Highest-Power Assembly; 100 hours after shutdown; all fuel pins damaged; all gap activity released; poor Iodine retention in Spent Fuel Pool Water. Noble Gas Release Iodine to Noble Gas Rate 0-2 hours.(Ci/S) _ Ratio Realistic Case 0.0117 0.00005 Conservative Case 2.233 0.000093 - Regulatory Guide 16.51 0.00073 1.25 Case NOTE: The realistic case values above are for one row of fuel rods. If more than one row is damaged, multiply the number of rows times the realistic case data. e-,m... . . - . .
EPP-005 ATTAC,HMENT XIX PAGE 7 OF *
/m REVISION 5 LOSS OF OFFSITE POWER ACCIDENT Accident Duration: 8 Hours Accident Assumptions:
Realistic Case: 100 lbs/ day Primary-to-Secondary Leakage; 0.12% Fuel -
~
. Defects prior to accident; 42 GPM Steam Generator Blowdown. Conservative. Case: 1.0 GPM Primary-to-Secondary Leakage; 1.0% Fuel Failure prior to Accident, 15 GPM Steam Generator Blowdown. 1 Noble Gas Release Iodine-to-Noble Rate (C1/S) Gas Ratio Realistic Case 0-2 hours 3.15E-6 O.00654 2-8 hours 3.15E-6 0.00362 Conservative. Case 0-2 hours 0.0218 1. 7 6 E'-3 2-8 hours 0.0218 8.54E-4 F s f p s
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. , 6 SOUTH CAROLINA ELECTRIC AND GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION
/~ '
NUCLEAR OPERATIONS s _.
\ LUC. EAR OPERAT ONS COPY No. . . . . . . . . . . /g7. . . . . . . . . .
EMERGENCY PLAN PROCEDURE EPP-010 PERSONNEL / VEHICLE DECONTAMINATION REVISION 3 NOVEMBER 3, 1983 SAFETY RELATED G // M
'Date DIS {fPLINESUPSRVISOR l A A h fA_i d /Ddte
//>/h APPROVAL AUTHORITY 9
)(OV DATE ISSUED RECORD OF CHANGES TYPE DATE- DATE CHANGE TYPE DATS DATE CHANGE NO. CHANGE APPROVED CANCELLED NO. CHANGE APPROVED CANCELLED ,
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_.._.....__..c_........__ . _ . . _ . . , _ _ . _. . . _ . _ . _ _. _ . . ..
., =s EPP-010 PAGE i REVISION 3 11/3/83 t
LIST OF EFFECTIVE PAGES PAGE REVISIO N i 3 , 1 11 2 - 1 2 2 2 3 3 4 2 i= ATTACHMENTS Attachment I . Deleted
, Attachment II. 2 ,
it
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I _.. _ . .
.. 6 EPP-010 REVISION 3 11/3/83*
- ~
4.1.6 If personnel decontamination is per5ormed at the offsite holding area, decontamination supplies located in the Q.A./ Security Building should-be utilized. 4.1.7 Complete Attachment I of HPP-405, " Personnel Contamination Report" for all contaminated personnel. , 4.1.8 Bioassays will be administered, based upon levels of skin contamination, at the discretion of the Radiological Assessment Supervisor. 4.2 Vehicle Decontamination. 4.2.1 Vehicle monitoring should be performed by Health Physics personnel using the appropriate equipment stored in the emergency kits. 4.2.2 The following limits will be used for unconditional , release of vehicles during offsite evacuation: : 4.2.2.1 Less than 1000 DPM/100 cm2 Beta / Gamma, smearable contamination. 4.2.2.2 Less than 100 cpm above background at 1/2" using ,
. - - a pancake GM Tube (HP-210 probe).
f
~'
4.2.3 All vehicles found to be contaminated in excess of the above limits should be decontaminated using the following guidelines. NOTE: Don appropriate protective clothing prior to decontamination attempts. 4.2.3.1 Rope and post an area accordingly. 4.2.3.2 Lay out plastic in the roped area and place the vehicle upon the plastic. 4.2.3.3 Using the decon supplies in the off-site monitoring kits, decon the vehicle using decon foam and towels or disposable rags. 4.2.3.4 Ensure that all contaminated towels and rags are placed in the radioactive trash bags provided in the Emergency Kit. PAGE 3 OF 4
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SOUTH. CAROLINA ELECTRIC AND GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION I NUCLEAR OPERATIONS hUCLEAR OPERAT ohs COPYNo.............l.51............. EMERGENCY PLAN PROCEDURE EPP-015 NATURAL EMERGENCY (EARTHQUAKE, TORNADO) REVISION 3 NOVLMBER 3, 1983 i SAFETY RELATED (.. ' Y A A. u , +w & # '7
' Bate
- DISCIPLINE SGPERVI3OR
. /YY> ta$.w APPROVAL AUTHORITY 1LI ,/ P2
' '6a tV N3V 9 983 DATE ISSUED RECORD OF CHANGES TYPE LATE DATE CHANGE TYPE DATE DATE CHANGE NO. CHANGE APPROVED CANCELLED NO. CHANGE APPROVED CANCELLED ,
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EPP-015 PAGE'i REVISION 3 11/3/83 LIST OF EFFECTIVE PAGES Pace Revision i 3 . 11 2 1 2 2 3 ATTACHMENTS Attachment I, Pages 1 through 4 2 \ e 4
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EPP-015 REVISION 3 11/3/83
- c. Designate Shift Personnel, or if the OSC is
"' activated, an Emergency Repair Team to ensure that all exterior doors to tho plant structures are shut and secure (include non-protected buildings if time permits). The Emergency. Repair Team should secure any equipment / materials in the Protected Area that could become missiles in high
- winds and put all heavy equipment in a safe condition (i.e., lowering crane booms).
I
- d. An Emergency Repair Team should stand by in the event emergency assessment and corrective actions e are required. The Emergency Repair Team should
' consist of the following:
- 1. Operations Personnel
- 2. Health Physics Personnel
- 3. Maintenance Personnel NOTE: Should a situation arise which would require the i presence of an armed individual such as work being done on the physical security system, assistance from the Security Force may be requested.
- e. If evacuation of.the site is required, follow the ,
w procedure cutline7 in EPP-012, Onsite Personnel Accountability and Evacuation.
- f. Upon cancellation of the watch / warning by the load dispatcher or the National Weather Service, make the appropriate announcement over the plant paging system and resume all normal activities.
NOTE: Backup to load dispatcher for weather information is the National Weather Service (Refer to EPP-002 for telephone number) 4.2 Earthquake 4.2.1 In the event of seismic disturbance, the Shift Supervisor will initiate the appropriate corrective actions for any resulting damage.
.4.2.2 If the seismic instrumentation in the Control Room indicates that the operating basis earthquake has been exceeded, the plant must be shutdown.
4.2.3 If evacuation of the site is required, follow the procedures outlined in EPP-012. . 4.2.4 For any seiemic disturbance, complete Attachment I and forward to Manager, Operations. N PAGE 2 OF 2 e
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SOUTH CAROLINA ELECTRIC AND GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION
, NUCLEAR OPERATIONS NUCEAR OPERATONS COPY No . . . . . . . . ./. 51. . . . . . . . .. ..
EMERGENCY PLAN PROCEDURE EPP-016 EMERGENCY FACILITIES ACTIVATION AND EVACUATION REVISION 4 , NOVEMBER 3, 1983 i SAFETY RELATED ( o // 'l f t DISCIPLINE SOFERVISOR / Date W ///>/ APPROVAL AUTHORITY Dat NOV 9 1983 DATE ISSUED RECORD OF CHANGES CHANGE TYPE DATE DATE CHANGE TYPE DATE DATE APPROVED CANCELLED NO. CHANGE APPROVED CANCI'LLED NO. CHANGE , n..t.- 1
~ ~". 1 :1'. .~ " _. _
EPP-016 PAGE i REVISIOff 4 11/3/83 LIST OF EFFECTIVE PAGES PAGE REVISION 1 4 11 3 1 3 2 3 3 4 4 4 5 3 3 I
- 6 7 3 ATTACHMENTS -
I . Attachment I 4 Attachment II 3 o C. l
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f_ .a . _ _ . _ - EPP-016 REVISION 4 11/3/83* 4.3.2 EOF / News Media Area per CEP-007. 4.4 At the discretion of the Interim Emergency Director, turnover of various responsibilities to the appropriate designated personnel can be accomplished as those persons respond to the emergency prior to the activation of the TSC/OSC. _ 5.0 PROCEDURE
~5.1 Activation of Technical Support Center (TSC) 5.1.1 Upon declaration of an Alert,. Site Emergency, or General Emorgency the following essential personnel shall proceed to the TSC.
A. Emergency Director (ED) B. TSC Communicator C. ED Administrative Assistants (2) D. Radiological Assessment Supervisor (1)
-. E. Out-of-Plant Monitoring Director F. In-plant Monitoring Director G. Personnel with:
- 1. Core Expertise
- 2. Mechanical Expertise i 3. Electrical Expertise 5.1.2 Upon declaration of an Alert, Site Emergency, or General Emergency the following key personnel should' proceed to the TSC.
A. Associate Manager, Security B. Mar.ager, Support Services C. Group Manager, Technical and Support Scrvices D. Deputy Director, Operations and Maintenance , k e a e PAGE 3 OF 7
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( EPP-016 REVISION 4 11/3/83
, E. Director, Nuclear plant Operations
~- l 5.1.3 At the discretion of the Radiological Assessment Supervisor, a radiological survey is performed to determine the habitability of the TSC. If the radiation dose rate is greater than 100MR/hr. or the airborne radiation levels exceed 10CFR20 Appendix B, Table 1 limits, s
the evacuation requirements of Section 5.2 of this procedure shall be carried out. 5.1.4 A personnel accountability check shall be performed. 5.1.5 The Emergency Director (ED) shall receive an update on the emergency Director (IED). condition from the Interim Emergency 5.1.6 Following the update of the ED, the IED shall officially turnover the Emergency Director duties and responsibilities to the ED along with the Emergency Log. 5.1.7 The emergency following the official log shall be kept by the ED's logger turnover. 5.1.8 The TSC communicator shall report to the CR to
,--. receive official turnover of the Emergency Notification Log from the interim communicator.
5.1.9 Upon the request of the ED, the TSC staff chall provide technical evaluations of the emergency condition and/or courses effects of the emergency of action to be taken to minimize the condition. 5.2 Evacuation of the TSC If the airborne radiation levels exceed 10CFR20, Appendix B, Table TSC, the 1 limits or theactions following radiation levels shall exceed 100 MR/hr. in the be taken. 5.2.1 The Emergency Director (ED) shall designate the personnel required to stay in the TSC and Control Room. , 5.2.2 The remainder of the personnel shall proceed to the Emergency Operations Facility (EOF). 5.2.3 If the EOF is or becomes uninhabitable designate the alternate assembly area., the ED shall -
,*e PAGE 4 OF 7 66M%*h"" *b _ -. .
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i ? - ATTACIIMENP 1 i PAGE 1 OF 1 . RhV. ISION 4 Brergency .
! Control
- I Officer l . t l l Ol'FSITE
- DIERGDU ____________________________,_______
MHEDC CDORDINNION DIRFC'IOR t i TIDINICAL CDtTI1OL l OPERATIONS EMBEENCY SUPPORT CENTER HDOM SGPIORT CDITER
! OPERATIONS FACILIT'? TEDINICAL RAD OPHMTIOt3 OPERATIONS SUP10RT ASSESSMENI SUPERVISOR SUPPORT CENTER ,
SIIIPT f 'ITDINICAL CDNS15"JCTION/ _ SUPERVISOR SJPERVISPI4 SUPBWISOR SUPPORT REPAIR PIRIP RAD SUPERVIfDR FIRE BRIGADE CDORDINNIOR (DORDINA'IOR SYSTHi ASSESSMEN1 OPHm10RS FIRST AID
& & ASSESSMENP PERSONNEL & OFHUTE STAFF STAFF PERSONNEL AUXILIARY OPES. SUPIT' t 3 MEDIA SEUJRITY R4MC JAIR l CDORDINNIOR (DORDINNIOR GENSEL OFFSITE RAD l
! SERVICES FDNI'IORItC l , (DDRDINNIOR CDORDINNIOR l EDF FEDHEL,STATF 6 ODNMUNICNIOli IDCAL FIGURE 5-5 OFFICIALS i IITTERFACE BE'lWEDI Oct. 1981 ONSITP/OFFSITE Rev. 5 H4DEDCY OPERATIOt3
SOUTH CAROLINA ELECTRIC AND GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION
! NUCLEAR OPERATIONS NUCLEAR 0?i3AT,0NS C0P.Y No........MZ EMERGENCY PLAN PROCEDURE .
EPP-017 POST-RECOVERY AND RE-ENTRY REVISION 4 NOVEMBER 3, 1983 SAFETY RELATED
,m
(.;
/2 Mi //l9 73 DIS PLINE.SCWERVISOR ' Bate sw //hlh ~
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APPROVAL AUTHORITY Date NOV 9 1983 DATE ISSUED RECORD OF CHANGES CHANGE TYPE DATE DATE CHANGE TYPE DATE DATE NO. CHANGE- APPROVED CANCELLED NO. CEANGE APPROVED CANCELLED .
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w EPP-017 PAGE i REVI'SION 4 11/3/83 LIST OF EFFECTIVE PAGES Page Revision i 4 11 3 1 3 2 3 ., 3 3 4 3 4 5 4 6 4 7 4
~
ATTACHMENTS f-s Attachment I 3 9 b 1.,
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a e w. _ , , m Mn EPP-017 REVISION 4 11/3'/83 A. Recovery and clean-up techniques are in accordance with good health physics practices and procedures. B. Personnel radiation exposure during the recovery 4 stage of the incident is closely controlled and documented. Individual exposures shall be in . accordance with administrative limits, where possible. C. Recovery activities and repairs are properly documented and controlled. D. Test programs to confirm fitness of equipment for return to service are developed and executed. 4.4.2 Based upon the extent of the emergency, long-term recovery operations may be required. The Emergency Director will notify the Offsite Emergency Coordinator and make recommendations as to the establishment of the Long Term Recovery Organization. 4.4.3 Recovery Oraanization(See Attachment I)
,-~.
A. The Of fsite Emergency Coordinator, with concurrence by the Emergency Control Officer and Emcrgency Director will shift from an Emergency Organization to a Recovery Organization. When a Recovery Organization is implemented, the Offsite Emergency Coordinator will inform appropriate offsite agencies. B. The Recovery Organization will depend upon the nature of the accident and the situation after the accident. A general organization is shown in Attachment I. Specific organizatio..a1 structure and staffing of the organization will be the responsibilty of the Emergency Control Officer. He will also determine the required scope and function of the recovery organization. The Emergency Control Officer is the Vice President, Nuclear Operations. C. The Recovery Manager, the Director, Nuclear l Services will be responsible for recovery ~ 8 activities, including technical decisions and
- courses of action. He will coordinate the functional activities of the various groups.
s 6 PAGE 5 OF 7 s
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EPP-017 REVISION 4 11/3/83 D. The Technical Support Manager, Group Manager, Technical Services is responsible for recovery activities including analysis and the development of plans and procedures in direct support of Plant Operations personnel with the objective of taking the plant to a safe shutdown condition in a manner which minimizes the effect on the health and safety
- of the public. This provides the central facility for the collection, retention, retrieval, and transmitting of plant and local environmental parameters.
E. The Plant Operations Manager, Director, Nuclear Plant Operations is responsible for implementation of in-plant recovery activities with the objective of taking the plant to a safe shutdown condition in a manner which minimizes the effect to the health l and safety of the public. F. The Construction / Repair Manager, Group Manager, Construction and Quality Control, coordinates the construction activities of the recovery operation. G. Advisory Support, a staff function consisting of senior technical personnel, will serve on the
~ Recovery Manager's staff. Typically, they consist
( , of senior representatives from the NSSS supplier, the A.E., and the NRC. Representatives for the recovery organization will be management personnel who have the requisite authority to
. represent their respective organizations in making resource commitments and resolving technical issues.
This group provides day-to-day support to the Recovery Manager and his staff and will be in attendance at the routine meetings of the recovery organization. H. The Scheduling / Planning Manager, Manager, Nuclear l Engineering serves as a support resource for the Recovery Manager, assisting him in the short-term planning, scheduling, and expediting of recovery operations. Usual measures are required to identify key problems; significant operations, technical decisions, resource limitations, and schedule milestones. The organization of this function will be strongly dependent upon the specific needs of the Recovery Manager. This group would covelop the agenda for recovery staff meetings and follow up
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with expediting of commitments made at these meetings. , PAGE 6 OF 7
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.. .- 4 EPP-017 REVISION 4 i
11/3/83
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I. The Administration and Logistics Manager, Manager, l Purchasing, provides administration, logistic, communications, and personnel support for the recovery operation. J. The Media Coordinator, Manager, Nuclear Information at the'V. C. Summer Nuclear Station, is the ranking ~ public information representative at the News Media Area who (1) relays information to other employees at the center, (2) supervises all communications operations at the center, (3) distributes all news releases and statements, and (4) maintains direct contact with public information specialists to ' develop guidance on policy and background material. The Media Coordinator will also coordinate - information at the news center with counterparts ' from local, state, federal agencies, and other companies involved with the emergency, and provide a means of meeting the media's needs. N) f b E'~ PAGE 7 OF 7
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SOUTH CAROLINA ELECTRIC AND GAS COMPANY VIRGIL C. SUMMER NUCLEAR STATION NUCLEAR OPERATIONS KUCLEAR OPERATLONS COPY No . . . . ! .2.i. . . . . . . .: . . EMERGENCY PLAN pROCLURE EPP-018 EMERGENCY TRAINING AND DRILLS REVISION 3 NOVEMBER 3, 1983 SAFETY RELATED
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DISCIPLINE SUPERVISOR ~ 'Date 4 4A ll fD APPROVAL AUTHORITY [ ate ' NOV 10 583
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DATE ISSUED RECORD OF CHANGES CHANGE TYPE DATE DATE CHANGE TYPE DATE DATE NO. CHANGE A" PROVED CANCELLED NO. CHANGE APPROVED CANCELLED , g 4S
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EPF-018 PAGE i REVISION'3 s 11/3/83 LIST OF EFFECTIVE PAGES PAGE REVISIO N i 3 11 2 + 1 2 2 3 3 3 4 3 5 3 6 2 ATTACHMENTS
- r. Attachment I, Pages 1 through 3 2 i
Attachment II, Pages 1 through 4 2 k N
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EPP-018 REVISION 3 11/3/83 *
-4.1.1 The Nuclear Education and Training Group is l responsible for coordinating the training of all Virgil C. Summer Nuclear Station personnel.
4.1.2 The Nuclear Education and Training Group may j delegate the training responsibilities to those personnel qualified to teach a specific subject. , 4.1.3 The instructor for Emergency Plan Training should - have an excellent understanding of the VCSNS Radiation Emergency Plan, Emergency Plan Procedures, and State and County emergency plans. Competence in instructional technique shull be verified through evaluation conducted by th? Nuclear Educatica and Training Group. The instructor shell be approved
' by the Director, Nuclear Plant Operations.
4.1.4 Lesson plans should be provided for each instructor's program and the designated instructor , will be responsible for identifying and correcting : weak areas within his part of the training program. 4.1.5 All on-site and off-site Virgil C. Summer Nuclear Station staff listed in the Training Groups and Associated Training Requirements (Attachment I) , shall be required to attend, on an annual basis, a 7- training session covering a review of the Emergency (,
"' Plan and Emergency Plan Procedures.
4.1.6 All non-essentialpersonnel, including contract people,_shall receive Emergency Plan training, on an annual basis, as part of the General' Employee ' Training Program, AP-1101. 4.1.7 All off-site emergency organizations listed shall be invited, on an annual basis, to attend an Emergency Plan and Emergency Plan Procedure Training Program per the requirements of Attachment I.
- 4.1.8 . Tests may be administered to attendees as deemed applicable by the instructor. These tests may be written or oral. A minimum grade of 70% shall be considered a satisfactory performance.
4.2 Emergency Plan Drill and Exercise Development b,
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EPF -018 REVISION 3 11/3/83 e4.2.1 The Emergency Coordinator ir responsible for the
- planning, scheduling and coordination of all emergency planning related drills and exercises, with assistance provided by other groups and individuals, such as the Nuclear Education and i I
Training Group. 4.2.2 All drills and exercises are approved by the Director, Nuclear Plant Operations, with the l'- exception of the Annual Radiation Emergency DIlll which shall be approved by the Vice President, - Nuclear Operations. 4.2.3 2ne following is the sequence of events for the development and implementation of drills and exercises: A. Determine the type of drill to be held and the date and time that it will occur. . B. Develop the scenario for the drill, including plant status prior to the occurrence of the emergency. NOTE: Scenarios for drills and exercises shall include but not be limited to, the s (w. following:
- l. -The bacic objective (s) of each drill and exercise.
- 2. The date(s), time period, place (s) 4 ar.d participating organizations,
- 3. The simulated events.
- 4. A time schedule of real and simulated initiating events.
- 5. A narrative summary describing the conduct of the exercises or drills to include such things as simulated casualties, offsite fire department assistance, rescue of personnel, use of protective clothing, deployment of radiological monitoring teams, and public information activities.
- 6. Arrangements for qualified observers.
PAGE 3 OF 6 l
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EPP-018 REVISION 3 11/3/83 , Drill scenarios shall be prepared that involve particication of several emergency teams and all cr cpecific parts of the on-site and off-site emergency organizations including varying degrees of participation of State, County, and Federal organizations and agencies, and local services support personnel and organizations. The Emergency . Coordinator shall notify the off-site emergency response organization and agencies at least thirty (30) days in advance of the scheduled date of the drill or exercise. C. Forward the completed scenario to the Director, Nuclear Plant Operations and the Vice President, Nuclear Operations for review and approval. D. Assign personnel who will be evaluators / controllers / obcurvers during the - course of the drill. E. A pre-exercise meeting will be held to brief evaluators / controllers / observers on the scope, sequence of events, and individual responsibilities.
< F. Conduct the drill as scheduled.
G. Upon completion of the drill, conduct a post-exercise critique during which time evaluators / controllers / observers comments shall be heard. The criteria for evaluation of the exercise is given in Attachment II. H. The names of all emergency personnel who participated in the drill shall be documented as per AP-1101 " General Employee Training." I. Recommendations of improvements in the Emergency Plan based upon the critique, shall be forwarded to the Coordinator, Emergency Planning for initiation of review and approval. Recommendations of improvements in the Emergency Plan Procedures, based upon the critique, shall be forwarded to the Director, Nuclear Plant Operations for his review and approval. J. All pertinent docuzents generated as a result of the drill shall be forwarded to the Emergency - Coordinator. PAGE 4 OF 6
-t ?t, EPP-018 REVISION 3 11/3/83*
<- K. Ensure that deficiencies discovered are reported
- , to the Director, Nuclear Plant Operations for review and corrective action.
4.2.4 Scheduled drills and exercises shall be held involving off-site as well as on-site emergency personnel, organizations, and agencies. , 4.2.5 ~Each scheduled drill will be conducted in a manner that ensures: A. The drill participants are familiar with their respective duties and responsibilities. B. The adequacy of the meth'ods used in the Emergency Plan Procedures. C. A check has been performed on the aval,1 ability of emergency supplies and equipment. , D. The operability _of emergency equipment. E. Plant conditions and operations will not be jeopardized. 4.2.6 The Emergency Coordinator is responsible for (,_' preparing drill scenarios. Managers shall provide assistance as requested by Section-Supervisors / i I the Emergency Coordinator to ensure the drills achieve the desired effect to the maximum extent possible. 4.2.7 The Emergency Coordinator should ensure the notification of.any offsite agencies to be involved in a drill at least thirty days prior to the' drill initiation. He should, as a reminder, also verify that the off-site agencies can still support their involvement in the drill by contacting them within forty-eight hours of the drill initiation. 4.2.8 Prior to the running of a drill, a pre-drill briefing should be conducted by the Drill Coordinator. All drill monitors should be in attendance to discuss the expected sequence of avents and to receive instructions on precautions to be observed during the drill sequence. After completion of the drill, the individual specified ao the lead Controller should conduct a post-drill '
. critique with drill monitors and participants at the earliest time when all applicable personnel can be gathered.
4.3 Energency. Plsn Drill and Exereire Schedule PAGE 5 OF 6
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