ML20136C664
| ML20136C664 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 07/24/1992 |
| From: | NRC COMMISSION (OCM) |
| To: | |
| Shared Package | |
| ML17229A261 | List:
|
| References | |
| FOIA-96-485 NUDOCS 9703120147 | |
| Download: ML20136C664 (400) | |
Text
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l SAINT LUCIE 1 VOLUME I- EMERGENCY RESPONSE N 31 7 970306
, BINDER 96-485 PDR w - _ _ -_ - - -. -
j l l I i SAINT LUCIE UNIT 1 . l I PLANT INFORMATION BOOK l l ( i s l PREPARED BY: University i of Maryland Student Researchers ' i
- EDITED BY
- John MacKinnon, ABOD
" ranCY REEPONSE SECTION BYs , NRR DRAWINGS PROVIDED BYt RFVIENED_BY1 '
ASSEMBLED BY: Joe Sebrosky, AEOD ,
. REVISION O O
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- NRC PLANT INFORMATION BOOK l
i SAINT LUCIE UNIT 1 i l TABLE of CONTENTS i i A. Facility Statistics l i B. Emergency Response information i , C. Plant Description Summary . l ' D. Simplified Plant System Diagrams
- E. Detailed Plant Systems Data i
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[, _ A. . d B. EMERGENCY RESPONSE INFORMATION
, SAINT LUCIE UNITS 1 and 2 -
i l TABLE of CONTENTS Emergency Response Facilities Site and Population Emergency Response Officials 1 Appendix: Drawings, Charts and Maps 1 i e 4 W (
. i NRC Plant Information Book
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- ' B. EMERGENCY PNONSE INFORMATION s ,
Plant Name: St. Lucie Unit 1 & 2 v EMERGENCY RESPONSE FACH2 TIES ' ! TSC: - ' Technical Support Center is located adjacent to the Unit 1 Control Room at i the St. Lucie Plant. Reference Attachment 1.
- . EOF: Emergency Operations Facility is located at the Midway Substation l approximately 10.1 miles due west of the power plant at the intersection of i state Road 712 and I-95. '
+ (
Reference:
10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) OSC: Operational Support Center is located in the North Service Building in the Conference Room on the second floor. (
Reference:
Attach =*=* 1) CORPORATE: Corporate OfEce is located in Juno Beach, Florida approximately 40 miles - h south of the plant on U.S.1.- . (
Reference:
50 mile EPZ Environmental Monitoring Map) State EOC: State of Florida Emergency Operations Center is located in Tallahassee, , Florida.- he State Warning Point (the 24 hour contact point for the Division of Emergency Management) is collocated with the EOC. Cou'aty EOC: He St. Lucie County Emergency Operations Center is located approximately 15 miles north west of the plant just off of Orange Ave. He St. Lucie County EOC is a. hardened facility per FEMA guidelines. (
Reference:
10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) He Martin County Emergency Operations Center is located approximately 15 miles south of the plant just off of Port Salerno Rd. The Martin County EO is a hardened facility per FEMA guidelines. (
Reference:
10 mile EPZ Environmental Monitori.ng Map for the St. Lucie Plant) e 0
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July 24,1992 d SITE & POPUL4 TION Site location: St. Lucie site is located on Hutchinson Island, FL Approximately 71/2 miles south of Ft. Pierce, FL (10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) Coordinates: LAT 20:20'55" N (Unit 1) LONG 80:14'47" W (Unit 1) LAT 27:20'58" N (Unit 2) LONG 80:14'48" W (Unit 2) (
Reference:
FSAR) Description of St. Lucie is a coastal plant (barrier island) with an eastern boundary of the the Environs: Atlantic Ocean and a western boundary of the Indian River. Population Distribution (Max population, resident and transient): 2-Mile Ring 658 5-Mile Ring 18,846 10 Mile Ring 124,019 Source: HMM d' ocument No 23%1/ENV/706 Imcal Conununities within the 10 mile EPZ: Ft. Pierce Port St. Lucie Stuart l West Palm Beach l I States within the 50 mile EPZ: 1 Florida i l Neamst Population Center (greater than 25,000): Ft. Pierce (5-10 miles) l Port St. Lucie (3-13 miles) j Source: HMM document No 23%1/ENV/706 , 2 Saint Lucie Units 1 & 2, REV 0
July 24,1992 EMERGENCY RESPONSE OFFICIALS Licensee' Representative with Authority to Make Pmtective Action Recommendations: Prior to staffing the Emergency Response Organization: Nuclear Plant Supervisor on shift in the emergency position of Emergency Coordinator. ! l '. Upon arrival in the Control Room and after proper turnover, the Plant Manager (or l other senior management member) can take over as the Emergency Coordirator. After Staffing the Emergency Response Organization: , j
- The Plant Manager (or other senior management member) can take over as the
- Emergency Coordinator in the Control Room or the TSC.
i When the EOF is operational and a proper turnover is given, the Recovery Manager assumes the responsibility for protective action recommendations from the Emergency Coordinator onsite. He also assumes offsite (state and local) notifications. ' l j l
- Licensee Representative with Authority to Make In-Plant Technical Recommendations
l l ) Nuclear Plant Superviser on shift in the Control Room functioning as the Emergency l Coordinator. i 1 Upon arrival in the Control Room and after proper turnover, the Plant Manager (or other senior management member) can take over as the Emergency Coordinator. After Staffing the Emergency Response Organization: j The Plant Manager (or other senior management member) functioning as the Emergency Coordinator in the Control Room or the TSC. , , When the EOF is operational, the Recovery Manager is the senior manager in charge of the emergency onsite. State / Local Government Representative with Authority to Make Protective Action Decisions:
- I Initial Stages prior to Governor signing an executive order and State and locals in their own EOC (not in EOF):
i i The highest ranking member of county management in the EOC (preferably the Chairmen of County Cornminion) with input from the State and DHRS Office of Radiation Control (Orlando). a i ' 3 Saint laicie Units 1 & 2, REV 0
9 _ July 24,1992 After the Governor signs an executive order and State and locals in their own EOC (not in EOF): The Governors Authorized Representative in the State EOC with input from the State DHRS Office of Radiation Control and local government. Prior to Governor signing an executive order and State and locals are in the EOF:
'Ibe county emergency management director in the EOF with' input from the State GAR and DHRS Office of Radiation Control.
After the Governor signs an executive order and the State and locals are in the EOF: The GAR in the EOF with' input from DHRS Office of Radiation Control and local government. l J 3 3
. 1 1
1 4 Saint Lucie Units 1 & 2, REV 0
SECTION B . j APPENDlX: RESPONSE DRAWINGS, CHARTS AND MAPS l ) i TABLE of CONTENTS 1 l Site Plan , 4 Emergency Planning Zone Sector Map . i
- 1 l Evacuation Routes )
i i Licensee Emergency Organization State Protective Action Decision Matrix
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i 1 i ! t NRC Plant !aformation Book i
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; COORDINATOR i-a
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i SHIFT ' TECHNICAL ASSISTANT ! ADVISOR NUCLEAR PLANT SUPERVISOR -- I i INTERIM NUCLEAR WATCH LICENSED PLANT ! RADIATION TEAM ENGINEER OPERATORS d LEADER ' L- 1 ^ NON-LICENSED l FIRE PMNT TEAM LEADER OPERATORS 4 ! INTERIM FIRSTAID/ DECONTAMINATION TEAM LEADER ' l INTERIM' SECURITY l TEAM LEADER (EPLNE 4.WP) l 4 d i 1
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- 4 2-19 St. Lucie
. Rev.26 1
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ATTAcunant 4c
~ FIEURE 2-1 '
FPL EMEREENCY RESPONSE OR2ANIZATION I EMERGENCY EMERGENCY '
- ' CONTROL INFORMATION '
OFFICER MANAGER lUj ' RECOVERY EMERGENCY GOVERNMENTAL M" MANAGER ' SECURITY - AFFAIRS ' ' LL - MANAGER MANAGER IL
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EMERGENCY 4
- TECHNICAL MANAGER ,
m- ., i l a NUCLEAR EMERGENCY PLANT GENERAL PLANT COORDINATOR MANAGER OR
- SUPV. (EC) ALTERNATE INTERIM TEAM LEADERS EMERGENCY TEAMS PRIMARY TEAM LEADERS h %
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- Z O Senior Health Ppysics Nealth Physke Radiation Supervisor orAllemato .
!. NuclearWatch Engineer Nucisar Watch Engineer !. Anemens (NWE) Fire Nucisar P W. (NPO) '; Senior Nuclear Plant Operator or First Aid & Chemistry Supervisor Associate Nuclear Plant Decontamination orAltomete Operator (ANPO) Security Shift Supervisor Plant Security Supervisor Altomate Security or l Protected Area Guard Ahemato W (Formed as Required) , (EPLN21.WPG) r i-d d EP3:4 - 2-4 St. Lucie Rev.26 4
,. ~ TABLE 2-2b FLORIDA POWER & LIGHT EMEREENCY RESPONSE OR2ANIZATION.
L- , FUNCTIONS AND RESPONSIBILITIES RESPONSIBILIT/ Function _ Immediate Exoanded .
- Command and Control Emergency Coordinator Recovery Manager (Nuclear Plant Supervisor)
Waming Emergency Coordinator Recovery Manager Notification _ Emergency Coordinator Recovery Manager Communications Public Information Emergency Information Emergency information Manager Manager
~ Accident Assessment Emergency Coordinator Recovery Manager (assisted (assisted by Shift Technical by Emergency Technical Advisor) Manager and his/her staff).
Fire Fire Team Leader Fire Team Leader Rescue Interim Radiation Team Primary Radiation Team Leader Leader Traffic Control Interim Security Team Primary Security Team (on-site) Leader Leader , Emergency Medical Interim First Primary First i Services Aid / Decontamination Team Aid / Decontamination Team l Leader Leader ! Transportation Interim Security Team Emergency Security , Leader
~
Manager l p < Protective Response Emergency Coordinator Radiation Team Leader (on-site) (assisted by RM's staff) Radiological Exposure Emergency Coordinator Radiation Team Leader
, Control _ (on-site) (assisted by RM's staff)
Radiological Dose Emergency Coordinator Recovery Manager (assisted I Assessment (assisted by Chemistry by Recovery Manager's Department representative) Staff)
. EP3:4 2-29 St. Lucie Rev.26 I
ATTACHMEAT 4e-FI' LURE 1-2 INITIAL N3TIFICATION EMERGENCY COORDINATOR Y STATE OF FLORIDA DMSION OF Ar DUTY CALL PLANTGENERAL EMERGENCY 4 SUPEFMSOR MANAGER UANAGEMENT - - Y STATE OF FLORIDA DHRS OFFICE E EMERGENCY L EMERGENCY . OF RADUmON CONTaa.
-k TEAM LEADER 7 TEAM MEMBERS EPZ COUNTIES NUCLEAR DMSION RE DIRECTORS DUTYOFFICER
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( k FPLOFFSrTE EMERGENCY r WWNN USNRC . SYSTEM OPERATION ' OPERATIONS % POWER COOR3NATOR l CENTER !
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LOCAL SUPPORT l INTERIM
- 1. FIRE /AMBULANCEJ EMERGE C
- 2. MEDCAL ' F MOEND l (1) V e Bases Not Rlrw Down Totaphone 01RD)
PRIMMtY NOT1RCATION PATHWAY g) Vna Emergency NadAcadon Syneem (ENS) ALTERNATE NOT1RCATION PATHWAY p) 60sdeel & Mre Emergencess orey, as needed m visan m ue sy=rn m (5) NDOO is the Emergency Connd Omoor for
- Nemmen EP3:4 1-14 St. Lucie Rev.26 1
3 ,g F12URE 2-5 EXPANDED RESPONSE ORGANIZATION e EMERGENCY CONTROL OFFICER EMERGENCY GOVERNMENTAL RECOVERY INFORMATON A M RS
. OFFICER MANAGER MANAGER I
I EMERGENCY EMERGENCY SECURITY TECHNICAL
, MANAGER -
MANAGER 4 1 i __
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ATTACHME NT 4 A e Fl2URE 2-2A STATE, LOCAL, AND FEDERAL RESPONSE BEFORE EXECUTIVE ORDER Attorney General Govemor Uceasse Department of Community Affairs l Risk Counties DMsion of Emergency Mana0ement Legend Direcison Host Counties OtherIngestion Exposure Pathway Counties Departmentof Health Department d Depanment Red # , and Rehabilitative Envimnmental Cross of 1.aw Services Regulation Enforcement Transpoen d Department of 1 Highway Safety & Natural Resources Motor Vehicles l l l l Depar%. . _ _ , ,,,.1 Depanment of Military Affairs A0riculture & Game & Fresh Water Consumer Seh Fish Commission (DFT2.WPG) t EP3:4 2-5 St. Lucie Rev.26 4
. l 1
Arremer Gs FIGURE 2-23 STATE, LOCAL, AND FEDERAL RESPONSE AFTER EXECUTIVE ORDER Attomey General Governor uoensee Department of Community Affairs l Division of Emergency Management j l Risk Counties i Host Counties . . Legend Direction Nlen Exposure Pathway Coordination Counties Department I Departmentof Health Department of Rg ntd and Rehabilitatrve Environmental Cross WW Tran'- w
-w^' n-Services Regulation Enforcement Department of Department of Highway Safety &
NaturalResources l Motor Vehicles .I Department of partmem d A0riculture & Game & Fresh Water Military Affairs Consumer Soh Fish Commission 4 (DFT3.WPG) i-l
- EP3:4 2-6 '
St. Lucie
. Rev.26
v, . t l l~ l i l i i i !I STATE PROTECTIVE ACTION i DECISION MATRIX t l f e i . . i e i ) I i. -b ) 4 s - I f 1 d .
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1 If . FIGURE R-3 PRIMARY AND. SECONDARY RESPONSIBILITIES ST. LUCIE COUNTY i-t . M E 5 ; I Si i2 g l s < O-H l 9 ss a s x - j
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- . LAW ENFORCEMEmi P
- TRANSPORTATION 8 I 1
FOOD OUALITY 8 P l POTABLE WATER OUALITY 8 P 8EELTER/ CARE- 8
- PUBLIC EEALTM AND SANITATION P SOCIAL SERVICES
- ROAD PASSAGE AND MAINTENANCE 1
SEuumITY P ~ TRAFFIC CGhruOL P 5 RECOVERY AND REENTRY 8 8 P i i *
, R-57 revision 12/31/s2 J
FIGURE R-5 i PRIMARY AND SECONDARY RESPONSIBILITIES l NARTIN COUNTY ' i i 1 v , 4-k E
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- PROTECTIVE RESPONSE P P S PUBLIC RLERT EMn NOTIFICATION P S 8 l PUBLIC INFORMATION P S S BADIOLOGICAL EXPOSURE CONTROL P j DECONTAMTMATION P 8 8 8 i CONTROL OF ACCESS TO EVACUATED mum 3 y g FIELD MONITORTMG 1un SAMDLTMG S P S i FIRE OPERATIONS P EMERGENCY MEDICAL OPERATIONS P LAW ENFORCEMENT P
, TRANSPORTATION P FOOD OUALITY P S POTABLE WATER OUALITY S P SHELTER / CARE 8 P S
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, TRAFFIC CONTROL - P RECOVERY AND REENTRY 8 8 8 8 P i-R-59 revision 12/31/92
M FIGURE R-7 i i PRIMARY AND SECONDARY RESPONSIBILITIES
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DECONThurMATION P S S CONTROL OP ACCESS TO EVACUATED FBW1 FIELD MONITORIMG mun 81usLING S P PIRE AMD RESCUE P-wwunGENCY MEDICAL SERVICES 'S P YA W ENPORCRMWMT P , TRAMSPORTATION P S S I FOOD OUALITY S P , POTAELE WATER OUALITY S P q BRELTER/ CARE P S S S S
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- RECOVERY AND REENTRY P S
. R-61 revision 12/31/92 4
l i l . l l* l 1. l FIGURE R-9 i PRInnRY AmD SECONDARY RIsPousIsILITIIs ! PALM BENCE COUNTY , l l l ! l 1 s i l 1 m E J { 5 5 i E
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- POTABLE WATER OUALITY P P SHELTER / CARE S P S
- PUBLIC EEALTE AND 8ANITATION P SOCIAL SERVICES P l
ROAD PAS 8 AGE AND MAINTENANCE P SECUmITY P TRAFFIC COurmOL P RECOVERY AND REENTRY 8 P R-65 revision 12/31/92 1
s f hLF h i B. EMERGENCY RESPONSE INFORMATl0N SAINT LUCIE UNITS 1 and 2 TABLE of CONTENTS l Emergency Response Facilities l Site and Population i i Emergency Response Officials - Appendix: Drawings, Charts and Maps ! l j l e 9 NRC Plant Information Book
l , i s ' i . B. EMERGENCY RESPONSE INFORMATION Y , , Pirnt Name: St. Lucie Unit 1 & 2 L EMERGENCY RESPONSE FACR2 TIES a TSC: Technical Support Center is located adjacent to the Unit 1 Control Room at the St. Lucie Plant. Reference Attachment 1.
. EOF: Emergency Operations Facility is located at the Midway Substation '
approximately 10.1 miles due west of the power plant at the intersection of - state Road 712 and I-95. (
Reference:
10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) ! OSC: Operational Sdpport Center'is located in the North Service Building in the i Conference Room on the second floor. (
Reference:
Attadunent 1) i i CORPORATE: Corporate Office is located in Juno Beach, Florida approximately 40 miles south of the plant on U.S.1. (
Reference:
50 mile EPZ Environmental Monitoring Map) State EOC: State of Florida Emergency Operations Center is located in Tallahassee, Florida. The State Warning Point (the 24 hour contact point for the Division of Emergency Management) is collocated with the EOC. County EOC: The St. Lucie County Emergency Operations Center is located approximately 15 miles north west of the plant just off of Orange Ave. The St. Lucie County EOC is a hardened facility per FEMA guidelines. (
Reference:
10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) The Martin County Emergency Operations Center is located approximately 15 miles south of the plant just off of Port Salerno Rd. The Martin County EO is a hardened facility per FEMA guidelines. (
Reference:
10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) l l I 1 h
. . . - . . , ,.v-- -
w " '
July 24,1992
, SITE & POPULATION l
1 Site Location: St. Lucie site is located on Hutchinson Island, FL. Approximately 71/2 miles south of Ft. Pierce, FL (10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) l i l Coordinates: LAT 20:20'55" N (Unit 1) - LONG 80:14'47" W (Unit 1) LAT 27:20'58" N (Unit 2) LONG 80:14'48" W (Unit 2) (
Reference:
FSAR) Description of St. Lucie is a coastal plant (barrier island) with an eastern boundary of the ] the Environs: Atlantic Ocean and a western boundary of the Indian River. ( Population Distribution (Max population, resident and transient): 2-Mile Ring 658 5-Mile Ring 18,846 10-Mile Ring 124,019 Source: HMM document No 2336-1/ENV/706 Local Conununities within the 10 mile EPZ: Ft. Pierce Port St. Lucie Stuart i West Palm Beach , States within the 50 mile EPZ: Florida ' Nearest Population Center (greater than 25,000): Ft. Pierce (5-10 miles) Port St. Lucie (3-13 miles) Source: HMM document No 23361/ENV/706 i i e I 2 Saint Lucie Units 1 & 2, REV 0
- -- -- . - - - . . . . ~ . . __. _ - ..- -. . - .- .. -
t July 24,1992
~ EMERGENCY RESPONSE OFFICIALS
- j. r l Licensee Representative with Authority to Make Protective Action Recommendations:
Prior to staffing the Emergency Response Organization: Nuclear Plant Supervisor on sitift in the emergency position of Emergency Coordinator. ' Upcen arrival in the Control Room and after proper turnover, the Plant Manager (or other senior management member) can take over as the Emergency Coordinator. After Staffing the Emergency Response Organization: ' The Plant Manager (or other senior management member) can take over as the Emergency Coordinator in the Control Room or the TSC. s I When the EOF is operational and a proper turnover is given, the Recovery Manager assumes the responsibility for protective action recommendations from the Emergency l i Coordinator onsite. He also assumes offsite (state and local) notifications. i i , Licensee Representative with Authority to Make In-Plant Technical Reconunendations: ; 1, Nuclear Plant Supervisor on shift in the Control Room functioning as the Emergency Coordinator. . i Upon arrival in the Control Room and after proper turnover, the Plant Manager (or other senior management member) can take over as the Emergency Coordinator. After. Staffing the Emergency Response Organization:
' Die Plant Manager (or other senior management member) functioning as the Emergency Coordinator in the Control Room or the TSC.
f. When the EOF is operational, the Recovery Manager is the senior manager in charge of
; the emergency onsite.
State / local G(vernment Representative with Authority to Make Protective Action
! Decisions: '
i ! Initial Stages prior to Governor signing an executive order and State and locals in their,own 4 EOC (not in EOF): The highest ranking member of county managem'" in the EOC (preferably the Chairmen of County Commission) with input.from the State and DHRS Office of Radiation Control (Orlando). i 3 Saint Lucie Units 1 & 2. REV 0
July 24,1992
' ' After the Governor signs an executive order and State and locals in their own EOC (not in EOF):- .
The Governorr, Authorized R'epresentative in the State EOC with input from the State DHRS Office of Radiation Control and local government.
; Prior to Governor signing an executive order and State and locals are in the EOF:
i '; The county emergency management director in the EOF with input from the State GAR and DHRS Office of Radiation Control. After the Governor signs an executive order and the State and locals are in the EOF: 1 The GAR in the EOF with input from DHRS Office of Radiation Control and local government. ' 2 l 4 l d I 1 4 1 h 4 Saint Lycie Units 1 & 2, REV 0
)
1 l i SECTION B APPENDIX: RESPONSE DRAWINGS. CHARTS AND MAPS 4 4
)
. TABLE of CONTENTS Site Plan ! i i
- Emergency Planning Zone Sector Map Evacuation Routes Licensee Emergency Organization State Protective Action Decision Matrix i
i i i i f i i t 1 1 NRC Plant Information Book
-- - - - - - . - - - o 9 i
l I ( t l f P l 1 SITE PLAN j e 4 6 M* 1 I l l
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- i
! ST. LUCIE PLANT EMERGENCY PLAN IMPLEMENTING PROCEDURE NO. 3100026E, REVISION 15 1
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- Page 13 of 13 ST. LUCIE PLANT .
EMERGENCY PLAN IMPLEMENTING PROCEDURE NO. 3100026E, REVISION 15
. CRITERIA FOR AND CONDUCT OF EVACUATIONS FIGURE 2 g o \ JAYCEE PARK
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'kYubd Y A 1 FIGURE 2-3 ST.
LUcTE PLANT NORMAL OPERATING ORCANTZATTON Vice Premdent 4 l Plant Manager Human Resource Plant SG . g Manager Engineer F%nt hos Manager I 1 Mh Technmal Dept. Opera $ons
*EE- ='W e 1 S %~.^4^ Tranng 95=. 441 9_ =, ,g&q 1
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IMMrnIATE RESPONSE ORGANT2ATION I EMERGENCY 1 C00f0NATOR 4 4 4 sHFT __ TECHPOCAL I N AssesTANT pywy NUCLEAR UCENSCD'
- - SUPUMoon - WATCH puwr NUCLE)W ENGNEER OPERATOR j NTERRA RADIATION TEAM LEADER 1
i FIRE j TEAM LEADER i 1 NTERM ' FRST AEM DECONTAAmeATION TEAM LEADER NTERR4
$ECURTY TEAM LEADER T
4 l . s a me l I EP3:4 2-17 s St. Lucia Rev. 21
g &L,nFW~' t i rrcune 2-s r
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i c v i EMEI43ENCY ,
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~ ~ EN N I INFOplMATION GOVEFINMENTAE N
AFFAIFB' - j omCER MANAGER i EMERGENCY EMERIENCY i SECURTY l TECHNICAL 4 MANAGER MANAGER i i . l . l l e EMER3ENCY ) COQfWINATOR - STAPPAs PWQUWBD l j - 4 i i I i . d i i. EP3:4 2-29 St. Lucia Rev. 21
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i
./Q / Nib & y g TABLE 2-2b Florida Power & Llaht ?mareanev Rannonna creanizatien Functions and Rammensibilitian Rammensibility Function 1332ft1&&A Exngnded Command and' Control Emergency coordinator Recovery Manager (Nuclear Plant Supervisor)
Warning Emergency Coordinator Recovery Manager Notification Emergency Coordinator Recovery Manager communications .
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Public Information Emerryency Information Emergency Info.
, Manager Manager Accident Emergency Coordinator Assessment Recovery Manager (assisted by Shift (assisted by Technical Advisor) Emergency Tech-nical Manager and his staff)
Fire Fire Team Leader Fire Team Leader Rescue Interim Radiation Team Prbaary Radiation Leader Team Leader I Traffic Control Interim Security Team Primary Security (on-site) Leader Team Leader l Emergency Medical l- Services Interim Firt Aid / Primary First Aid ; Decontamination Team / Decontamination Leader Team Leader Transportation Interim Security Emergency Team Leader. Security Manager Protective Response Emergency Coordinator Radiation Team (on-site) Leader (assisted by RM's staff). ! Radiological !' Emergency coordinator Radiation Team Exposure Control (on-site) Leader (assisted by RM's staff) Radiological Dose . Emergency Coordinator Assessment Recovery Manager
* (assisted by Chemistry Department representative) 1
]
l EP3:4 2-25 St. Lucie Rev. 21
M MM Page 8 of 24 k ST. LUCIE PLANT i5 l E-PLAN IMPLEMENTING PROCEDURE NO. 3100023E, REVIS! ON SITE EMERGENCY ORGANIZATION AND CALL MGURE 2 INmAL NOT1MCATION_M 1
- l i
j EMERGENCY COORDINATOR , I i
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N INFORMADON GOVERNMENTAE. N AFFAlm - OFFICER 4 MANAGER 1 i EMElWRENCy ! EMERGENCY ssCURmf i- MANa g TECHNICAL b MANAGER l EMERGENCY ADmflONAL COOftl1NATOR STAFFAS REQuuuD l 1 i 1 4 . 4 , 1 EP3:4 l 3 2-29 St. Lucia ' Rev. 21
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STATE, LOCAL, AND FEDERAL RESPONSE BEFORE EXECUTIVE ORDER 1 I 4. ATT W N '****************- m ' --- ... < UCENGEE 4 4 :
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I DEPMme!NT OF COMMUMTY AFFAMS ! 1 Rl8K COUNTES OM800N OF EnERGENCY MANAGEnENT
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. . . .. _ . . .- . = . . . -. . . .- .. -. .. -.
l l l k FIGURE 2-2b . STATE,' LOCAL, AND FEDERAL RESPONSE AFTER EXECUTIVE ORDER i { mof.EY 00E,M . ................ o0ve 0R f............... l uCEN.EE, 5 i 1 1 DEPARTnENT OF CCAmaMTY AFFAfte DMes0N OF EMERGENCY MANAGEMENT 4 ! l i ' 140KCOLWTES l
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' ; 1 ! UEP W ' DEPARTnENT OF MGHWAY SAFETY '****'
DEPARTnENTOF DEPARTnENTOF Afe MOTCft VENCtK8 NATUfMLPESCURCES i ) MUTARYAFFAMS AGfeCLA.TupEAec CONeuhER SOMCE8
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1 EP3:4 2-6 St. Lucia Rev. 21
1 STATE PROTECTIVE ACTIOX DECISION MATRIX J Y i' I f , e l i - I l J 4 f 4 4 4 t i s lM l 4 e
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l , 1: 1 l 1 4 . l 4
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- l .
Je C. PLANT DESCRIPTION
SUMMARY
l-4 SAINT LUCIE UNIT 1
; .e 1
NOTE: This Section was reproduced from Chapter 1 of the Final Safety l 4 Analysis Report (FSAR) l 4 1 i i 1 i i I l
. 1 l
NRC Plant infennation Book l
i t i i ! 1.2
~
CENERAL PLANT DESCRIPTION l < '. 1.2.1 PRINCIPAL SITE C11ARACTERISTICS The site'for the St Lucie Plant consists of approximately 1,132 acres. . The unimproved area of the site.is generally flat, covered with water and. has a ' dense vegetation characteristic of Florida coastal mangrove' swamps. At the ocean shore the land rises slightly in a dune or ridge to approximately 15 j feet above mean low water. i The island and the adjoining mainland are sparsely populated. The nearest j population center is the City of Fort Pierce which .is eight miles from the site and has 29,721 people as of the 1970 census.- The minista site exclusion j radius is 5,100 feet. Site characteristics are given in detail in Chapter 2.
- 1. 2. 2 .1 Principal Desian Criteria j
Principal structures, systems and equipment which may serve either to prevent accidents or to mitigate their consequences are designed and are erected in accordance with applicsble codes to withstand the most severe earthquakes, i i flooding conditions, windstorms, temperature and other deleterious natural phenomena which could be reasonably assumed to occur at the site during the , lifetime of the plant. Principal structures, systems and equipment'are sized 4 for the design power level of the nuclear steam supply system output. i Redundancy is provided in the reactor protective and safety feature systems so j , that no single failure of any active component of the systems can prevent 4 action necessary to avoid an unsafe condition. The plant is designed to l facilitate inspection and testing of systems and components whose reliability
- are important to the protection of the public and plant personnel.
Provisions are made to minimize the probability and effect of fires and { explosions. i Systems and components which are significant from the standpoint of nuclear ' safety are designed, fabricated and erected to quality standards commensurate i with the safety function to be performed. i
)
Section 31 addresses the implementation of the NRC General Design Criteria for Nuclear Power Plants,10 CFR Part 50, Appendix A. Chapter 17 describes j the Quality Assurance Program. d 3 1.2.2.2 Reac tor i
; The following design criteria apply to the reactor:
4 a). The reactor is of the pressurized water-type, designed to provide heat ' j to steam generators which, in turn, provide steam to drive a tur,bine generato r. The initial full power c' ore thermal output is 2560 megawatts. 4 4 en -
, f 8 y 1.2-1 . +
2
- b) The reactor fuel is slightly enriched uranium dioxide contained 'in Zircaloy tubes.
c) . Minimum departure f rom nucleate boiling ratio (DNBR) during normal operation and anticipated transients is not below that value which could lead to fuel rod f ailure. The maximum center line temperature of the fuel, ' evaluated at the design overpower condition, is below that value which could lead to fuel rod failure. The melting point of the . UO2 is not reached during routine operation and anticipated transients. d) Fuel rod clad is designed to maintain cladding integrity throughout fuel life. Fission gas release within the rods and other factors affecting design life is considered for the maximum expected exposures, i e) The reactor and control systems are designed so that any xenon transients will be adequately damped. . f) The reacto'r is designed to accommodate safely and without fuel i I l 1 1.2-2 i I 1 1
. . - - -~ ~ . . - - . . - - . - . . . . ~..- .- --
n --
+ , , j damage, che' antic'ipated operational occurrences,' backed up by auto- !
matic and redundant reactor trips, t i , g) The reactor coolant system is designed and constructed ,co, maintain
-its integrity throughout the plant J~s e. . Appropriate means of
~
test and inspection are provided. - 1 h) Power excursions which could result from any credible reactivi.cy addition will not cause damage, either by deformation or rupture l of the reactor vessel and will not impair operation of the engine- . j ered safety features. > I , i ' i) Control element assemblies (CEA's) are'. capable of holding the core suberitical at het zero power conditions with margin following a-trip, . even' with the most reactive CEA stuck in the fully with- .; y . drawn position. i 1 ] j) The chemical and volume control system is capable of adding boric { i acid to the reactor coolant at a rate sufficient to maintain an , adequate shutdown margin during reactor coolant system cooldown at ' i the maximum design rate following a reactor trip. The system is l- independent of the CEA system. i i k) The combined response of the fuel = temperature coefficient, the . moderator temperature coefficient the moderator void coefficient and the moderator pressure coefficient to an increase-in reactor i thermal power is a decrease in reactivity. In addition, the reactor power transient remains bounded and damped in response i to any expected changes in any operating variable. > Further details concerning the reactor are given in Chapter 4. 1
, 1.2.2.3 Reactor Coolant and Auxiliary Systems j/ .
' Heat removal systems are provided which can safely accommodate core heat
- output. Each.of the heat removal systems is designed to provide reliable '
operation under all normal and expected transiest circumstan'ces. (See
- chapters 5 and 9). '
i Components of the reactor coolant system are designed and vill be operated so that no excessive pressure or thermal stress will be impose / on the i structural materials. The necessary consideration has been given to the ductile characteristics of the materials at low temperatures. 1.2.2.4 Containment Structure The containment structure, including the access openings and penetrations, > ] _ is designed to contain the pressures and temperatures resulting from a . i l ': loss of coolant accident (LOCA) in which: (a) a range of reactor coolant pipe breaks, up to a' double-ended break of the largest reactor coolant ' {f ,
- 1. 2 -3 ,
I
a l pip 3, is assumid end the govsrning snargy relsase into the containmant l therefrom is calculated; (b) there is a simultaneous loss of external I electric power; (c) additional heat is transferred from the reactor to con-tainment by water supplied from the safety injection system; (d) either the containment cooling system functions or one half of the containment spray system functions; and (e) the engineered safety features are actuated 30 . seconds following an accident, except that the safety injection tanks oper- i ate when the system pressure drops below the tank pressure. I i i '; ~ Means are provided for pressure and leak rate testing of the entire con- l tainment system including provisions for leak rate testing of individual i piping and electrical penetrations that rely on gasketed seals, sealing ' compounds, or expansion bellows. Integrity. of the containment is assured against postulated missiles from plant equipment failures and against postulated missiles from external l sources. (See Chapters 3 and 6 for details). , j
- 1.2.2.5 Engineered Safety Features The plant design incorporates redundant engineered safety features. These '
systems in conjunction with the containment system, ensure that the offsite l radiological consequences following any LOCA up to and including a double ended break of the largest reactor coolant pipe, will not exceed the guide-lines of 10CFR100. The systems ensure that the Final Acceptance Criteria for. Emergency Core Cooling Systems for Light Water Power Reactors (Federal Register Vol 39, Jan. 4, 1974) are satisfied, based upon analytical methods, assumptions and procedures accepted by the NRC. The engineered' safety fea-l tures include: (a) independent redundant systems (containment cooling system and containment spray system) to remove heat from and reduce the pressure in 3 the containment vessel in order to maintain containment integrity, (b) a high and low pressure safety injection system to limit fuel and cladding damage to an amount which will not interfere with adequate. emergency core cooling, and to limit metal-water reactions to negligible amounts, (c) a shield building ventilation system to reduce offsite consequences due to leakage from the containment vessel; (d) a containment isolation system to minimize l post-LOCA radiological effects offsite, (e) a hydrogen control system to
, maintain safe post-LOCA hydrogen concentration within the containment.
(f) control room habitability system, and (g) an Iodine Removal System to remove radioactive' iodine from the containment atmosphere to minimize post-LOCA affluents. (See Chapter 6 for details). f 1.2.2.6 Instrumentation and Control i Interlocks and automatic protective systems are provided along with adminis- . trative controls to ensure safe operation of the plant. A reactor protective system is provided which initiates reactor trip if.the reactor approaches prescribed safety limits.
- Suf ficient redundancy is installed to permit' periodic testing of the reactor protective system so that failure or removal from service of any one protective system' component or portion of the system will not preclude reactor trip or
~
i other safety action when required. -
- 1. 2-4 ' -
1
i The protactive systcm is s2perstsd from the control instrumtntation systems so that failure or removal from service of any control instrumentation ; system component or channel does not inhibit the function of the protective s ys t em. (See Chapter 7 for details.)
-1.2.2.7 Electrical Systems Redundant sources of off-site power are provided by three separate transmission lines to ensure that no single failure of any active component can prevent a safe and orderly shutdown. !
, Redundant sources of emergency on-site power are provided by two diesel 4 generators to ensure safe shutdown in the event of complete loss of off- l site power. (See Chapter 8 for details.) 1.2.2.8 Waste Management and Radiation Protection l The waste maassement system is designed to provide controlled creatment , and disposal of liquid, gaseous, and solid wastes. The principal design ' criterion is that plant personnel and the general public are protected by ; ensuring that a11' normal operating releases of radioactive material are ' made as low as practical in accordance with the provisions of 10CFR50, Appendix I. (See Chapter 11 for details.) l l ' Units 1 and 2 are provided with separate control rooms,' housed in their ' 1 respective reactor auxiliary buildings, each having adequate shielding to permit occupancy during all postulated accidents involving radiation releases. = The radiation shielding in the plant, in combination with plant radiation control procedures, ensures that operating personnel do not receive radi-ation exposures in excess of the applicable limits of 10CFR20 during normal operation and maintenance. (See Chapter 12 for details.) 4 1.2.2.9 Fuel Handling and Storate New and spent fuel handling and storage facilities (Section 9.1) are pro-i vided for the safe handling, storage and shipment of fuel and are designed to preclude accidental criticality. ; l 1.2.2.10 Power Conversion f The steam and power conversion system is provided to generate steam in i two steam generators for energy transfer co a regenerative cycle turbine-generator which generates electricity. After expanding in the turbine, . the steam and is to returned condensed the steamin the main and generators condenser. The condensate is collected reject heat ! is dissipated in the circulating water system. (See Chapter 10 for details.) 1.2.3 OPERATING CHARACTERISTICS AND SAFETY CONSIDERAT10NS 1.2.3.1 Nuclear Steam Supply System
~ l The reactor core is fueled with uranium dioxide pellets enclosed in Zircaloy tubes with welded and plugs. The tubes are fabricated into assemblies in which end fittings prevent axial motion and grids prevent i
1.2-5
j . . . i lateral motion of the tubes. The control element assemblies'(CEA's) j
' consist . of Inconel clad ~ boron carbide absorber ' rods which are guided by Zircaloy tubes located within the fuel assembly. The core consists of 217 fuel assemblies loaded with three different U-235 enrichments.
2 i t The reactor vessel and its closure head are' fabricated from manganese-moly steel internally clad with stainless steel. The vessel and its internals are designed so that the integrated neutron flux (greater
- athan 1.0 Mev) 40-year period. at the vessel wall will be less than 1.91 x 1019 nvt over i J
j
- - The internal structures include the core support barrel, the core sup-
- port. plate, the core shroud, the thermal shield, and the upper guide .
, s tructure assembly. - The core support barrel is a right circular cylinder 1 ' supported from a ring flange from a ledge on the reactor vessel. The j
flange carries the entire weight of the core. The core support place . transmits the weight of the core to the core support barrel by means of. vertical columns and a beam structure. The core shroud surrounds the ~ core and minimizes the amount of coolant bypass flow. The upper guide
]
structure of the fuelprovides assemblies a flowduringshroud fo,r the CEA's pressure and prevents upward motion , transients. Lateral motion limiters ~ i or snubbers are provided at the lower and of ~ the core support barrel assembly,
)
.The reactor parallel coolant to the reactor system is arranged as two closed loops connected in vessel. Each loop consists of one 42-inch ID
' outlet (hot) pipe, one steam generator, two 30-inch ID inlet (cold)
. pipes and two pumps. An electrically heated pressurizer is connected j
j to the hot leg of one of the loops and a safety injection 14.ne is con-nected to each of the four cold legs.
- I i The reactor coolant system operates at a nominal pressure of 2235 psig.
[ The reactor coolant enters near the top of the reactor vessel, and flows-i- downward between the reactor vessel shell and the core support barrel-thermal shield into the lower plenum. It then flows upward through the j core, leaves the reactor vessel, and flows through the tube side of the , two vertical secondary U-tube steam generators where heat is transferred to the system. the reactor vessel. Reactor coolant pumps return the reactor coolant to 4
- The two steam generators are vertical shall and U-tube units. The steae generated in the shell side of the steam generator flows upward through i moisture separators and scrubber plate dryers which reduce the moisture content to less than 0.2 percent.
coolant are eithsr stainless steel orAllNiCrFe corrosion. surfaces in in alloy contact order with to minimize the reactor
; The reactor suction coolant vertical centrifugal is circulated pumps.by The four electric motor driven single-
- pump shaft leakage is minimized by mechanical. seals. Each pump motor is equipped with an antireverse mechanism to prevent reverse rotation of any pump that is not in operation.
1.2.3.2 Enkineered Safety Features and Emersency Systems Engineered safety features systems protect the public and plant personnel 1.2-6 . 8 e
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l in1the- highly un1' 1kely event of anLaccidenta15 release of radioactive fission products f rom the reactor. system,~ particularly as the result of v a LOCA. The safety features function to localize, control,' mitigate, and terminate such. accidents to hold exposure levels below applicable limits. .
- The engineered safety features are:
a) The safety injection system (including high pressure and low. I pressure safety injection pumps and the safety injection tanks) b) . The containment sys tem - i j i. c) The containment spray system i
- d) The containment cooling system i ,
- . e) The shield building ven'tilation system 4- ,
f)' s The containment isolation system I-1 13 ) The hydrogen control' system . h) The control room habitability system a In the event o'f a LOCA, the safety inj.ection system described in Section {- 6.3 injects borated water into the reactor coolant system. This provides ; i cooling to limit core' damage and fission product release,' and assures [ adequate shutdown margin. The injection systems also provide continuous
.long term post-accident cooling of the ' core by recirculation of borated
" water from the containment sump ~ chrough the shutdown heat exchangers and back to the reactor core. F ! The containment is comprised of a steel containment vessel surrounded by j a reinforced concrete shield ' building. ' The containment vessel is a low i' leakage steel shall which is designed to. confine the radioactive material that could be released from a postulated hypothetical accident resulting in substantial core meltdown and release of fission products as defined , in 10CFR50. 'It is a cylindrical vessel with hemispherical done and el-lipsoidal bottom: !. The shield building is a medium leakage concrete
- structure which surrounds the steel containment vessel. It protects 3
the containment vessel from external missiles, and provides biological shielding 'and a means of collecting radioactive fission products that
- i. may leak from the' containment following a major hypothetical accident (MHA).
(See Section 6.2.1 for details.) The containment in conjunction with either of the associated spray and cooling systems is designed to withstand the internal pressure and coin-
- cident temperature resulting from the energy release associated with the worst postulated LOCA at a core power level of 2700 Mwt.
.b The containment is equipped with two 100 percent capacity spray systems ^ and an ' independent full' capacity containman't cooling system for cooling j the containment atmosphere following the postulated LOCA. (See Section 4
. 6.2.2' for details.)- -
3 s 4 1.2-7
The containment sprays supply borated water to cool and reduce pressure l in the containment atmosphere. The system is designed so that with one l spray pump, one set of spray nozzles, and one shutdown cooling heat ex-changer in operation, adequate cooling is provided to cool .the containment
~
atmosphere. The pumps take suction initially from the refueling water t ank. Long term cooling is based on suction from ~ the containment sump through the recirculation lines.
.The containment cooling system is designed to provide containment at-mosphere mixing by recirculation. The cooling coils and fans of the i
containment cooling system are sized to provide adequate containment cooling at post-accident conditions of temperature, pressure and humidity. l The shield building ventilation system is 'provided to maintain a negative pressure in the annulus between the steel containment building and the s concrete shield building following a LOCA. Two independent 100 parcent i capacity systems are provided. This system filters any radioactivity leakage from the containment vessel and therefore reduces the effects } on the environment. (See Section 6.2.3 for details.) ! An isolation system consisting of valves and associated actuators and j controls is provided for each line penetrating the containment that must i be closed to prevent a radioactivity release in the case of a loss of
, coolant accident. (See Section 6.2.4 for details.) i A hydrogen control system is provided which consists of hydrogen recombiner, hydrogen purge and hydrogen sampling systems. The purge system is provided as a redundant and diverse backup to the recombiner system.
The control room habitability system is provided to limit control room doses.due to airborne activity to within GDC 19 limits and' permit person-nel occupancy during a chlorine release accident. The system is discussed in detail in Section 9.4.1. ! 4' i The iodine removal system is provided which is designed to operate in . conjunction with the containment spray system to remove radio-iodines from the containment atmosphere following a loss of coolant accident (LOCA). This greatly reduces the quantity of radio-iodines released to the environment in'the event of post-LOCA leakage from the containment vessel. d 4 1.2-8
1.2.3.3 Protection, Cont rol, Ins t rumentation and riect rical Systems a) Reactor Protection ' i l The reactor parameters are maintained within the acceptable limits by the ' inherent characteristics of the reactor, by the reactor regulating system, by boron in the moderator and by the operating procedures. In addition, in order to preclude unsafe conditions for plant equipment or personnel, j the meterreactor reachesprotective its presetsystem limit. initiates reactor trip if a selected para-l Tour independent channels normally monitor each of the selected plant parameters. The reactor protective system logic is designed to initiate protective action whenever the signal of any two of four channels reaches the preset limit. If any two of these four channels receives coincident signals, the pmeer supply to the magnetic Jack control element drive mechanism is interrupted releasing the control j elements to drop into the core to shut.down the reactor. Redundancy is provided in the reactor protective system to assure that no single failure will prevent protective action when it is required. The protective system 4 is completely independent of and separate from the control system. ; b). Control System The reactor'is controlled by a combination of control ele' ment assemblies (CEA's) and dissolved boric acid in the reactor coolant. Boric acid is used for reactivity changes associated with large but gradual changes in water temperature, core xenon, fuel burnup and power levels. Additions I
- of boric acid also provide an increased shutdown margin during the initial loading and subsequent refuelings. The boric acid solution is prepared and stored at a temperature sufficiently high to prevent precipitation.
CEA movement provides changes in reactivity .for shutdown or power changes. The vessel CEA's head.are actuated by control drive mechanisms mounted on the reactor The control drive mechanisms are designed to permit rapid J ' insertion of the CEA's into the reactor core by gravity. CEA motion can l be initiated manually or automat'ically. The reactor regulating system (Section 7) provides for startup and shut-down turbineof thedemand. load reactor and fo'r adjustment of the reactor power in response to The nuclear steam supply system is capable of fol - lowing a ramp change from 15 percent to 100 percent power at a race of 5 percent per minut, and at greater rates over smaller load change incre-ments up to a step change of 10 percent. This control is normally ac-complished by automatic movement of CEA's in response to a change in
.the reactor coclant automatic temperature, signal at any time.with manual control capable of overriding A temperature controller compares the existing average reactor coolant temperature with the value corresp'onding to the power called for by the temperature control program.
temperature is different, the CEA's are adjusted to bring the If the two tem-paratures within the prescribed control band. Regulation of the reactor coolant stesa pressure within operating limits and matches reactor p demand. t 9 I 4 1.2-9 '
l ' W3 prosauro in tha rsecter coolant system is centro 11sd by rsgulcting tha ' temperctura of tha crolant in tha pressurizer, where steam and water are held in thermal equilibrium. Steam is formed by the pressurizer hesters or condensed by the pressurizer spray to reduce variations caused by expansion and contraction of the reactor coolant due to reactor system temperature changes. We pressure
- and water level control systems are described in Section 7.7.1.2.
, Overpressure protection of the rescur coolant system is provided by operated relief valves and spring loaded safety valves connected to the
- pressuriser and designed in,accordance with Sectica III of the ASME code. In addition, overpressure protection of the reactor coolant system during low temperature, solid system mode (i.e., startup), is provided by the Overpressure
- Mitigation System described in Section 5.2.2.6. ' he discharge from the pres-suriser safety and relief valves is released under water in the pressurizer s
~ quench tank, where it is condensed and cooled. In the event the discharged steam exceeds the capacity of the tank, the tank relieves to the containment ' ] atmosphere. A turbine control system is provided to regulate steam flow to the turbine as 4 a function-of system load. In the event of turbine trip, bypass systems are i provided to release steam to the condenser and to the atmosphere. Rose systems are designed to reduce the sensible heat in'the reactor coolant system, main- , 3 tain the steam generator pressure during hot standby and permit turbine trip without opening the steam generator safety valves when the condenser is avail-1 able. ! l
- A steam generator wate'r level control system regulates feedwater flow to the
! steam generator. i (See Section 7.7.2.3) An auxiliary feedwater system is pro-vided to ensure flow to the steam generators in the event the main'feedwater l supply is out of service. j c) Instrumentation System i
%e nuclear instrumentation includes out-of-core and in-core neutron flux f detectors. Ten channels of out-of-core instrumentation monitor the neutron I j
flux and provide reactor protection and control signals during start-up and power operation. Four of the channels monitor the neutron flux through the 1 start-up range, and six channels monitor the neutron flux from within the start-up range through the full power range. Of the latter, four are used for reactor. protection and two for reactor control. (See section 7.5.2.) i i he in-core instrumentation consists of self-powered rhodium and vanadium neutron detsetors and thermocouples to provide inforination on neutron flux distribution and temperature in the core. a : 1 3
' he process instrumentation monitoring includes those critical channels which are used for protective action. Temperature, pressure, flow and liquid level l 4
monitoring is provided, as required, to keep the operating personnel informed i of plant conditions and to provide information from which plant processes can be evaluated and/or regulated. The baron concentration in the reactor coolant
] water is also monitored and the concentration is displayed in the control room.
Instrument signals penetrating the containment are electric. Instrument signal transmission for the remaining plant instruments is either electric j or pneumatic. (See Chapter 7 for details.) i 1.2-10 4
, - ~
4
' Th2 plent g:ssous' and liquid afflunnes are annitorad to casura that they are !
i* saintained within acceptable radioactivity limits. Activity levels are displayed ; and off-normal values are annunciated. Area.sonitoring stations seasure r radioactivity at selected locations in the plant for. personnel protection. A j complete description of the instrumentation is contained in Chapter 7. . i i' d) Electrical . Systes , j h plant includes a 1,000 ava, 0.85 power factor generator delivering power to a l j 240 kv switchyard through step-up power transformers. Auxiliary power is utilised at 6.9 kv, 4.16 ky, 480 v, and 120 y ac; 125 y de systems are also , l ' provided for emergency power, engineered safety features control, and essential j nuclear instrumentation. I > j i h auxiliary load is nores11y supplied from two auxiliary transformers connected to .the main generator bus. Start-up power is supplied from two start-up j 4 transformers connected to the 240 kv switchyard. Energency power for the engineered safety features is supplied by redundant diesel generator sets. (See j Chapter 8 for- details.) j 1 1.2.3.4 Power Conversion Systen e i h turbine generator is a Westinghouse Electric Corporation unit. It is an i 1,800 rpe tandeo-compound, four-flow exhaust unit with 44 inch last stage l blades. N feeduster pumps are electric mot.or driven. Each of two. strings of l j feedwater heaters consisto of four low pressure and one high pressure heater. ; 1 1 ! The auxiliary feedwater systes contains two 325 spe electric motor driven pumps and one 600 spe pump driven by a non condensing stese turbine. This system is designed to provide emergency heat removal capacity. (See Chapter 10 for j details.) ~ 1.2.3.5 Fuel Handlina and Storane Systems i The fuel handling systems provide for the safe handling of fuel assemblies and control element assemblies under all foreseeable conditions and for the required l- assembly, disassembly, and storage of the reactor vessel head and internals. These systems include a refueling eachine located inside containment above the refueling cavity, the fuel transfer carriage, the tilting eachines, the fuel i transfer tube, a spent fuel handling eachine in the fuel handitag building, and ' various devices used for handling the reactor vessel head and internals. (See Section 9.1.4 for details.) i 3 New fuel is stored dry in vertical racks in the fuel handling building. Room is l provided for storing one third of a core. The rack and fuel assembly spacing 4 precludes criticality. (See Section 9.1.1 for details.) i-The fuel pool is a reinforced concrete structure, stainless steel lined, which contains high density spent fuel storage racks consisting of individual cells ! with 8.65 inch by 8.65 inch (nominal) square cross-section. Each cell can accomodate a single Combustion Engineering or Exxon PWR fuel assembly or
- equivalent, from either St Lucie Unit 1 or Unit 2. A total of 1706 cells are arranged in 17 distinct modules of varying size in two regions. Region 1 is designed for storage of new fuel assemblies with enrichments up to 4.5 4
weight percent U-235. Region 1 is also designed to store fuel assemblies with enrichments up to 4.5 weight percent U-235 that have not achieved adequate burnup for Region ' 2. The Region 2 cells are cr.pable of accommodating fuel assemblies with' various initial , enrichments which have accumulated aisimus burnups with an ! acceptsble bound. ' i 0120F 1.2-11 As. 8-7/89
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.__m h
(( l 4 !* 1 i i i Cooling and purification equipment is provided for the fuel pool water. This equipment may also be used for cleanup of ref ueling water a,f ter each fuel change in the reactor. (See Section 9.1.3 for details.) L 1.2.3.6 Coolina Water and other Auxiliary Systems 1 a) Chemical and Volume Control System The purity level in the reactor coolant system.is controlled by continuous i purification of a bypass stream of re' actor coolant. Water removed from ' the reactor coolant system is cooled in the regenerative heat exchanger. From there, the coolant flows to the letdown heat exchanger and then j through a filter and a domineralizar where corrosion and fission products.
,are removed. It is then sprayed into the volume control tank and returned i to the regenerative heat exchanger by the charging pumps where it is heated
- prior to return to the reactor coolant system.
The chemical and volume control system automatically adjusts the amount of reactor coolant in order to maintain a constant level in the pressuriser. l This compensat'es for changes in specific volume due to coolant' temperature i ! ' changes and reactor coolant pump shaft controlled seal leakage. (See ' Section 9.3.4 for details.) ) i j b) Shutdown Cooling System i j - The shutdown ecoling system is used to reduce the temperature of the reactor i coolant at a controlled rate from 325F to a refueling temperature of approx-imately re fueling. 135y and to maintain the proper reactor coolant temperature during
- . The shutdown cooling system utilizes the low pressure safety injection i pumps to circulate the reactor coolant through two shutdown heat exchangers, i
returning it to the reactor coolant system through the low pressure in-jection header. (Section 9.3.5) 1-The component cooling system serves as a heat sink for the shutdown heat exchangers. { { c) Component Cooling System 1
- The component cooling system, consisting of three pumps and two heat ex-changers, removes heat from.the various auxiliary systema. Corrosion i.
inhibited domineralized water-is circulated by the system through all d components of the nuclear steam supply system that require cooling water. During reactor shutdown, component cooling water .is also circulated through the shutdown heat exchangers. . The ~ component coeling system provides an intermediate barrier between the reactor coolant system and the intake cooling water system. (See Section 9.2.2 for details) d) Sampling System i , i j- '.Two sampling systems are provided; one for the reactor coolant and its
^ auxiliary systems, and one for the turbine steam and feedwater system.
l ,! i o 1. 2-1R
.t i . Y These sys tems are 'used for determining both chemical and radiochemical j
' conditions of the various process fluids used in the plant. '
e) . Cooling Water Syste'as j .The turbine generator condenser is cooled by the circulating.watec system
, which takes suction from and discharges to the Atlantic Ocean. :
1-
' ' An intake cooling water system provides seawater from the circulating water system intake structure and serves as a heat sink for the component coolittg water heat exchangers, the turbine closed cooling system heat exchangers, j and the steam generator open blowdown heat exchangers.
4 The turbine closed cooling system removes heat from the turbine generator l .
! oil cooler, hydrogen coolers, feed pump oil coolers, sample coolers, and l other components by providing corrosion inhibited demineralized water to
- those components. l
} f) Plant Ventilation Systems ' Separate ventilation systems are provided for the contain' ment. vessel, the control room, the. reactor auxiliary .buildit.g the fuel handling building 4 and the diesel generators building. A purge system is provided for the i containment vessel atmosphere.
- The annular 5 shield buildi ce between the steel containment vessel and the concrete is ventilated by the shield building ventilation system utilizing charcoal filters.
ation following a postulated LOCA. This system is automatically, put into oper- { (See Section 9.4 for details.)
- 3) Plant Fire Protection System
, The fire protection system supplies water to fire hydrants, deluge syster.s and hose racks in the various areas of the plant. , t ! Noncombustible a'nd fire resistant materials are. used throughout the facil-ity, particularly in areas containing critical portions of the plant such as the containment' structure, control room, cable spreading room, and t rooms containing components of the engineered safety features systems. A number of portable fire extinguishers are placed at key locations for use in extinguishing limited fires. ! (See Section 9.5 for details.) L 1.2.3.7 Radioactive Waste Manaaement System i The waste management system provides the means for controlled handling,
- storage and disposal of liquid, gaseous and solid wastes.
In addition. , the system reconcentrates effluent for reuse in the plant. and recovers ~ dissolved baron from the liquid Reactor coolant from the chemical and volume centrol system and from the
- reactor drainoftank is processed by the boron processing system, which'is
- comprised filters, i and condensata tanks. flash tank, , ion exchangers, concentrators, holding boric acid. The concentrators are used to' reconcentrate the The concentrate'is normally returned to the boric acid 1.2-13 l
makeup cank in the chemical and volume control system, but if the solu- ! tion is unsuitable for reuse, provision is ma'de to concentrate and trans-port it from the plant to a disposal site. The distillate from the con- ' centrator is sampled and may be discharged to the circulating water system if the radioactivity is within specified limits. . Miscellaneous liquid wastes from the reactor auxiliary building are col-1ected in the equipment and ' chemical drain tanks and subsequently processed
' by filtration, ion exchange and concentration. The distillate enters the waste condensate tank. If the radioactivity level of the liquid in the condensate tank is found to be high, the waste can be recycled through )
i the waste ion exchanger and' waste concentrator. The liquid in the holding tank is sampled to ensure radioactivity levels are within the acceptable limits prior to discharge to the circulating water system. The concen-trate is drummed for off-site disposal. Waste gases are compressed and stored in the gas decay tanks which have a 30-day storage capacity. After decay, the gas in the waste gas decay tanks is sampled to ensure radio, activity levels are within acceptable limits, and is then released to the plant vent at a controlled rate. Spent ion exchange resins and filters are ultimately transported in a shielded container from the plant. , Low activity' wastes such as contaminated laundry, rags and paper are com-pacted and drummed for. removal f rom the plant. (See Chapter 11 for o= tails). 1.2.4 l MAJOR STRUCTURES AND EQUIPMENT ARRANGEMENT i The turbine building for the St. Lucie Plant is oriented parallel to State Road A1A and the shoreline of the Atlantic Ocean, with the re-actor containment structure located on the east, or seaward, side of the l turbine building. i The reactor auxiliary building is located perpendicular and close to the turbine building, oriented in an east-west direction.
' The fuel handling building is located next to the reactor containment building dir'ction.
e and the reactor auxiliary building, oriented in a north-south 1 I The service building is located north of the turbine building. Refer to the Site T1ot Plan, Figure 1.2-1, and the Plant Plot Plan, Figure 1.2-2, for the si'.e general layout. and sections are shown in Figures 1.2-3 through-1.2-19.The plant structures arrang The containment structure houses the nuclear steam supply system (SSSS), consisting of tne reactor, steam generators, reactor coolant pumps, pres-surizer, and some of the ether reactor auxiliaries. The containment structure is served by a polar bridge crane. The auxiliary building houses the waste management facilities, engineered safety features components, heating and ventilating system components, switchgear, laboratories, offices, laundry and control room. 4 d 1.2-14 i
l' f
' . The fuel handling building contains the spent fuel pool and new fuel f
storage facilities, as well as the cooling and purification equipment ' for the fuel pool. -The fuel is transferred from the reactor containment building to the fuel handling building through the fuel transfer tube. The turbine building houses the' turbine generator, condensers, feedwater ! heaters, condensate and feedwater pumps, turbine auxiliaries and certain of the switchgear assemblies. ! The service building provides offices, shop and warehouse space, and is located next to the turbine building unloading bay, l l 1 l 1 l e t 4 e 1.2-15' , Am. 4-7/86
i . { I r 1.2.6 SHARED SYSTEMS AND INTERCONNECTIONS BETWE.EN UNIT 1 AND UNIT 2 i Normal shutdown of Unit I would require the operation of the following } ' auxiliary systems: , a) Auxiliary Feed b) Camponent Cooling i c) Intaka Cooling { d) Instrument Air l , None of these systems are shared between Unit I and Unit 2. A tie , between the two units has been provided between the condensate storage j tanks, but this is normally closed and is operated only when safe operation of the two units can be assured. The elevation of this tie connection on the tanks assures that the minimum quantity of water required for safe shutdoun is maintained at all times in both tanks. 1 Section 3.1.5 discusses other systems, including the. instrument air j
+
supply, which are interconnected but not shared for normal operation. Flow diagrams show the interconnection points between the two units. As a means of minimizing the disruption of Unit 1 operation and/or safety, the tie connections will be piped in such a way that the inter- !
- connection can be achieved without affecting the operating system.
i i 1 Following completion of Unit 2 construction, the independence of the j units is maintained by keeping the systema isolated at ! i all times. The diesel oil system and the auxiliary feedwater/ condensate I
! storage tank crosstie are the only interconnecting tie lines; !
Class I in the case of the former and non-seismic in the case of the i j latter. Class I locked closed isolation valves assure that the tie
' . lines will be opened only after administrative approval has been ob-tained.
1-
; 1.2.6.1 Startup Transformers -
l l The start-up transformers are sized to accommodate the auxiliary loads of the unit under any operating or accident condition. Each set of j i start-up transformers (1A-2A,1B-25) is provided with a manual switching arrangement which permits paralleling 4.16 kv power to Units 1 and 2 under administrative. control. In the event one of the four start-up transformers l' has to be removed from service for repair, the 4.16 kv power to both Units j j
'1 and- 2 is paralleled to facilitate continued operation of both units. A
" single startup transformer is adequately sized to accommodate the auxiliary l 1.2-22 Am. 4-7/86 b
t I-
. loads of both units under accident conditions when aligned as described above. If it should be neces.sary to start a unit with the other unit operating and with one start-up transformer aligned to supply 4.16 kv power to both units, appropriate operating procedures will be developed to assure ~
y thac the start-up transformer is not overloaded should an accident condition arise. 4 1.2.6.2 STATION BIACKOUT CROSS-TIE In the event of a Station Blackout ( 880 ) on either St. Lucie Unit, the SBO tie is utilized to power the minimum hot standby loads of one AC power train of each unit until conclusion of the SBO event. 1.2.7 SYISOLS AND ABBREVIATIONS ON FIGURES Definitions of symbols and abbreviations used thre>@~2t the chapters on fluid and electripal systems are shown in detail.on Figures 1.2-20 through 1.2-30. p 4
)
b l i
, 1.2-23 d[NE5li,~Ti'2/93)- . .
i e l l i
~ ; D. SIMPLIFIED PLANT S STEM DIAGRAMS ; i FIGURES; DRAWINGS: CHARTS: AND USTS FOR PWR's i
This section contains simplified one-line " training" diagrams which are suitable for display to the. Executive Team during a technical briefing. l
- s. Roosest Coolant System ,
; b. Chemises and voksne Comros system (CvCs) l ; unakading serie Acid systemi 1 I
j c. '-- r,;rci Core Cooling System (ECCS) ! o High Head (HHI) o intermedises HeadilHI) l i o Low Head (LHO
, o Acoumuissers {
o Rooircuiselon Phase Alignment l
- o Any Other Mont Unique ECCS
- d. Residual Heat Removel(RHR) System
- e. Containment Cooang System i
- f. Ceneminment Sprey(CS) System j g. Aux 5ery Feedweser System (APW)
- h. Main Stoem System I. Condensees and Feedweser System
, J. Emergency Elecotood Disetbution System:
- o swhenverd
! o 4180/480 VAC , ! o DC System (250/125 VDC) o ineeument Power System (120 VAC) i ! k. Safety Releend 0:r;:w;;/Servios Cooling Water Systems
- 1. Reester Vessel and Imamel Drawinge l
- m. vessel Weenr Level Instrument Range Diagrams !
1
- n. Neumon Monhering System Range Diagrams i
. o. Containment Drawinge ! p. ECCS Pump Performance Curves ' o HHI e lHE '
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o M , o RHR l o APW *
- q. IlSecedoneous IEuseselve Drawings r
4 e s 4 4 e i e i 1 Reactor Coolant System l l i l l 6 l l 9 i e t 4 4
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' INTAKE' COOLING WATER SYSTEM FLOW'OlAGRAM i- FIGUR'd 1: -
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d 9 9 P E e a i l . Neutron Monitoring System Range Diagram e i i e 9O i 4 e l l l I e 9 e a a S e w -
.. . _ , . . . . . . - .. - - . .- . . . - -.. . - . . . . . _ . - . . . . . . - _ = - . . . - - . . -. . - - _ -
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- - - - - , _ , - _ _ , - - - - --_-____----_~-_-a_----,-- , - m -_-- u % a-_.
t 1 ECCS Pump Performance Curves j
. 1 i
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- NPSH, ,
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plEVISICpt 9804 PflOCEDuplE Tm.5; i 6- DAGE. APPENDIXES / FIGURES / TABLES
" * * " " """ 55 of 78 1 EOP-99 EMERGENCY OPERATING PROCEDURE ST. LUCIE UNIT 1 i
PIGURE 2 SAFETY INJECTION PLOW VS.'RCS PRESSURE ' IE
- ' iiiii I
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l i-- i PWR DRAWINGS 4 b i A. General Reactor Coolant System l i a q 8. Chemical and Volume Control System (including Boric Acid System) C. ECCS -- 4
-High Head
-Low Head
! -Accumulators
~
4
-Recirculation Phase Alignment / System ;
, D. Residual Heat Removal ^
- E. )
Containment Cooling (safety related) ' F. Containment Spray G. Auxiliary Feed Water H. Hain Steam I. Condensate and Feedwater System J. Emergency Electrical Distribution System
- . Switchyard :
.4160/480 VAC
.DC System (250/125 VDC) l . Instrument Power (120 VAC)
K. AllCoolingWaterSystemswhichinterfacewithSafetyRelated Systems I - i L. Reactor Vessel and Internals Drawings M.
- Vessel Water Level (RVLIS) Instrument Range Relative to Active Fuel 4 l
N. Neutron monitoring System Range Diagram ! O. Reactor Containment Drawing l P. All ECCS Pump Performance Curves L S e j
- I i
e .- PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E. DETAILED PLANT SYSTEMS DATA l PLANT NAME UNIT-SAINT LUCIE UNIT 1 i TABLE of CONTENTS
~
, 1. Facility Statisties
- 2. Reactor Coolant System l
; 3. ESF and Reactor Trip Technical Specification Serpoints
- 4. Chemical Volume and Control System (CYCS) l
- 5. N/A .
i l
- 6. Intermediate Head (Safety) Injection System (IHI) j 7. N/A
- l
- 8. N/A
- 9. Low Head Injection (LHI)/ Residual Heat Removal (RHR) Systems
- 10. Safety Injection Tanks l l !!. Borated Water Sources for HHI, IHI, LHI, CVCS
- 12. Condensate, Feedwater, and Circulating Water Systems
! 13 Auxiliary Feedwater Supply Sources
- 14. Auxiliary Feedwater System
- 15. Main Steam System
- 16. Reactor Containment
- 17. Reactor Containment Cooler System
- 18. Containment Spray System
- 19. Alternate AC power Sources
- 20. Intake Cooling Water System
- 21. Component Cooling Water System s
. Appendix-A ' Abbreviations
- 4. 1 4
d e NRC Plant Information Brak 03/93
4 PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT I REVIEW SHEET Section A: Facility Statistics AND Section E: Detailed Plant Systems Data NAME WJ DATE ORIGINAL BY: REV-0 REVIEW REY.I REVIEW: l i REY.2 REVIEW: REV.3 REVIEW:
) ; REV-4 REVIEW:
REV 5 REVIEW: REV-6 REVIEW: REV-7 REVIEW: REY.s REVIEW: REV.9 REVIEW: REV.10 REVIEW: ' 4 NRC Plant Information Book 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 Elol l 1.0 FACILITY STATISTICS : i l i 1 Plant Name/ Unit: Saint Lucie 1
- i 1
- 2. Docket Number: 50-335 j 3. Owner / Operator: Florida Power & Light Co. F1.1 l
- 4. Other Plante On Site: Saint Lucie Unit 2 j 5. Site Location / Address: Hutchinson Island in Saint Lucis County between Fort F1.1 Pierce and Stuart.12 miles southeast of of Fort Pierce, j Flonda.
- 6. MAX Facility License Thermal 2700 F1.1
. Power Limit: (MWl)
{
- 7. Main Turbine Generator Rated 890 (grose)
' ~ ~~
f F1.1 j Output: (MWe) 4 =
. 8. Ultimate Heat Sink - Seppl/ to Atlantic Ocean F9.2.7.2 Main Turbine Condenses-l l
- 9. Ultimate Heat Sink - Supply to Atlanta Ocean F9.2.7.2 l ECCS Service Water System (s):
- l l 10. PWR Type: 2 Loop F9.2.1.2 1
;. 11. NSSS Vendor: Cornbustion Engineering F1.1
, 12. Turbine Generator: Westinghouse F1.2.3.4
- 13. Arch. Engineering: Ebesco Services, Inc.
i
- 14. Constructor: Florida Power & Light, Ebasco Semca F1.4
- 15. Date of Operating License 03/01/76 NRC Information Issuance
- Digest l 16. Date of Commercial Operation: 12/21/76
- NRC Informat6en i
Digest 1 a. NOTES: I i f J l I i , o , NRC Plant Inforsaation Book 03/93 . l
PRESSURIZED WATER REACTOR UNIT. SAINT LUCIE UNIT 1 E2-1 2.0 REACTOR COOLANT SYSTEM General
- 1. MAX T.S. RCS Preuure Safety 2750 T32.1.2 Limit (peig) :
4
- 2. Number of Tuel Anembliu : 217 TS$.3.1
- 3. Number of Fuel Roda per 176 / TS5J.1 Anembly :
Pressurizer (PZR)' Level
- ja I1 O[ / 'O I , ,,g[
- 4. PZR Hester Cutoff Level ,
Setpoint: (Units) *
- 5. T.S. MIN Operating PZR Level:
f b) 90v ____- (Unite) ".-
,7 - - , t. 14.
t - .
+L"g^_
I Pressurizer Power Operated Relief Valves (PORV)
- e. Valve Identification No. - V.1404 v.1402 FTb5.5 5 (Sheet 5.5 33)
- 7. Open Prueure Setpoint(s) : (paig) 2885 s/ sses t/ FTb5.5 5 /
lpfff g( (Sheet 5.5 as)
- 8. PORY operator : ko#lenb Iokno[d FTb5.5 5 (Sheet 5.5 33)
- 9. PORY Power Supply Bus : gggj k7 nu. voltese :g%) n5 12 ; rTb5.5 5 i
I j f \l b
.a
,rz , -
gi irv 9C suct ib fuc4 IB NRC Plant Information Book 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT I E2 2 - i Pressurizer PORY Overpressure Protection (NDTT) Setpoint I
- 10. OPEN Preneure Setpomie(s) s ,
~~
- s. Heatup : (peia) 75555[S e T53.4.13 Note 1)
- b. Cooldown :(peia)
M N b #)f (
- 11. OP9N'dfemperature N ....isi:
W a. Heatupf :(des f35h F) 0 ek ~ ~ ~ ~ ~
"j'< ~3*F/ T83 413 193'T &
- b. Cooldown : (des F) 5304'T(See Note 1) 15'F/ T53.4.13 5'F &
l'F
- 13. No. PORVs required by T.S.: 3 F5.3.3.6 & T53.4.13 (for NDTT protection with no i
other options synilable) l
- 13. T.S. MIN Vent Opening : (eg.in.) I
).75 T. s . D e F%, ., > . . n Io.! (
j (PORY NDTT alternative)
- 14. Other T.S. Options for NDTT (other than PORVs and vent openings)
Protection i PORY Block Valves (if applica'ble)
- 15. Valre Identification No. V-1405 V-teos FTb5.5-6 &
- Fas.1-3
- 16. Valve Power Supply Bus : MCC 1A-5 MCC 1B 5 FTb8.3-1 (pg 44,48)
Bue Voltage : (VAC) 480 4ao FTb8.3 1
' (pg 44,48) i i i
Pressurizer Code Safety Valves 1 17, Number of Safety Valves : 3 FTb5.5-4 l
- 18. T.S. Pressure Relief Setpoint : 3500:1% T33.4.3 (psia)
NRC Plant Information Book 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT I E2 3 Pressurizer Spr.cy Veins y
- Q
- 19. Valve Identification No. < 1100E 1100F FTb5.5 7 & -
Fs5.1 3 ,
- 20. RCS pump ID that supplies spray: IB2 1B1 FFgl.1-3 y
- 21. Auxiliary Pressuriser Spray Supply System Name:
CVCS F9.3.4.1 g q b b 5 Reactor Coolant Pumps (RCP)
- 22. Pump Identifleation No.: 1A1 1A2 1B1 IB2 FFs5.1 3
- 23. Pump Power Supply Bus : IB.1 1A 1 1B.1 1A.1 FFs5.1 3
,~
y Bus Voltage : (VAC) 6.9KV 6.9KV 6.9KV 6.9KV FTb8.3 1 f3 Q- I , Reactor Coolant Pump (RCP) Motor Cooling 419 s, NN
- 24. System *A* Name : .%
Component Cooling, System F9.2.2.2 g
- 25. System 'B' Name : . g
- 26. Both Systems Required :
s (i.e.. Dat redtme%:t) ]
% 4 Reactor Cociant Pump (RCP) Seal Inje los
%g %y As
- 27. Syetem "A* Name : CVCS F9.3.4.2.1 4
., N
- 28. System *B* Name : gylgf** Y
~
- 29. Both Systems Required : (i.e.. D21 redund ut)
D Reactor Coolant Pump (RCP) Thermal Barrier Cooling sg %, c.
- 30. System "A" Name : Component Cooling System FFg5.5-7
(
- 31. System "B" Name : gg' M.
- 32. Both Systems Required : ' (i.e., noi redundant)
*s NOTES:
1. () Tech spec lista different setpoints for relief protection bened on Reactor Coolant System Temperature. The values are as fallows: 350 pain for heatup'and temp sj93'T M - 530 psia for heatup and temp,>,,193*F & 5304*F
! . 350 pois for cooldown and temp <215'T
/ . 530 psia for cooldown and temp 15'T & <281'F M'e 2,'
NRC Plant Qf'ct-In(ormatihtikooQqjyfQ'*M k hv l# rs ty,wr 03/93 dm*
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 . E3 1 3.0 ESF ACTUATION TECHNICAL SPECIFICATION SETPOINTS i l i t i 1 A i - i i J 4 2 Y NRC Plant Information Book 03/93 ________________e _ _ _ _ _ _ _
4 TABLE 2.2-1 E REACTOR PROTECTIVE INSTRUN NTATION TRIP SETPOINT LIMITS O FUNCTIONAL UNIT m _ TRIP SETPOINT
- 1. . ALLOWABLE VALUES ~
Manual Reactor Trip Not Applicable z , Hot _ Applicable q , 2. Power Level - High (1) '
- Four Reactor Coolant Pumps Operating 1 9.615 above THERMAL POWER, with a minimum setpoint of 15% 1 9.61% above THERMAL POWER, and of RATED. THERMAL POWER, and a a minimum setpoint of 15% cf RATED maximum of < 107.0% of RATED THERMAL-POWER and a maximum of THERMAL P06fR. -
< 107.0% of RATED THERMAL POWER 3.
~ Reactor Coolant Flow - Low (1)
.i Four Reactor Coolant Pumps -
Operating > 95% of design reactor coolant
> 95% of design reactor coolant T10w with 4 pumps operating *
- 4. T10w with 4 pumps operating
- Pressurizer Pressure - High -1 2400 psia 1 400 2 psia
- 5. Containment. Pressure - High '
1 3.3.psig 1 3.3 ps19 6.
.Y Steam Generator Pressure - Low (2) 1 600 psia 1 600 psia
$ 7.
Steam Generator Water Level -Low g 1 20.5% Water Level - each 1 19.5% Water Level - each
- s steam generator steam generator l
- 8. Local Power Density - High (3) if Trip'setpoint adjusted to not Trip. set point adjusted to not exceed the limit lines of Figures 2.2-1 and 2.2-2 exceed the limit lines of Figures 2.2-1 and 2.2-2.
co
- Design reactor coolant flow with 4 pumps operating is-370.000 gpm. -
5 .a l
-_------._--.--_._-_a--.,..__.______--__1-- -
_ . . _ _ m. . 1 I TABLE 2.2-1 (Continued) U l . g REACTOR PRDTECTIVE INSTRUENTATION TRIP SETPOINT LIMITS -l i Q rn FU1CTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES
- 9. Thermal Margin / Low Pressure (1)
=
y - Four Reactor Coolant Pumps Trip setpoint adjusted to not Trip setpoint adjusted to not-
. Operating exceed the limit ifnes of Figures 2.2-3-and 2.2-4. exceed the Ilmit lines of i Figures 2.2-3 and 2.2-4.
9a. Steam Generator Pressure 1 135 psid 1 135 psid DifferenceHigh(1)
.(logic in TH/LP)
- 10. Loss of Turbine -- Hydraulic >806psig- > 800 psig FluidPressure-Low (3) -
t
$ 11. Rate of Change of Power - High (4) 1 2.49 decades per minute 1 2.49 decades per minute TABLE NDTATION ,
(1) Trip may be bypassed below 11 of RATED THERMAL POWER; bypass shall be automatically removed when-THERMAL POWER is 1 1% of RATED THERMAL POWER. ' h (2) Trip may be manually bypassed below 685 psig; bypass shall be automatically removed at or above i g 605 psig. ' g (3) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed
~ when THERMAL POWER is > 15% of RATED THERMAL POWER.
" i (4) Trip may be bypassed below 10-4% and above 15% of RATED THERMAL . POWER.
w b i i e o ____._._____m__ . _ _ _ _m.m ___._ . _ _ . _ . _ _ _ _ . _ _ _ _ _ __ _ . _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - .
- i. .
. r.. . . . . .
1 4 1 ! 1.0
,, _ _ ;;,;,;. _ _ . _ ,_,;; ,=
a i i GA 1 F - . 1 w # M i { ' m 1 t l u l. I 0.2 t 0 ! 0 0,2 OA
- 0.8 OA 1.0 FRACTION OF RATED THERMAL POWER t
- FIGURE 2.21 Local Power Density - High Trip Setpoint l
Part 1 (Fraction of RATED THERMAL POWER Versus 2 OR I t s . I i 1
- ST. LUCIE - UNIT 1 2-6 ,
t . i .. . l I
- d i ;
h d i i
1.4 - E 1.3 - UNACCIFTA2 2 UCN g g 3y . (0.0,1.17) OPERATION 1.1 - ( .2,1.0) 9, ( . 2 ,1. 0 ) 0.s - E 0.6 - 0.7 - 3(=.6,0.66) l O.6 - 9 (.6,0.66) 3 ' 0.5 - b ACCEPTABM GPERATION - 9 0.4 - j OJ - ~ 02 - . 4 O.1 - l j 0 y , , , , , , , ,
-0.7 l
-0.5 - 0.3 -0.1 0.1 0.3 C.6 0.7 AXIAL SHAPE INDEX, Y1 i
i d
. PIGURE 2.2-2 I4CAI. DOWER DENSITY- IIGE TRIP SETPOINT PART 2 (QE2 Versus T1)
ST. LUCIE - UMIT 1 27 Amendment NO. 27.32.88.106
. W 0 s Ian 1
_ - - - - J
4 i
' 1.3 .'Al FUNCTION 4
j , i Fygg = 2061 Al ,
- 1.4 . QR1 + 15.85 TIN - 8950 1 - .
d
- 1. 3 '
1 0 1 J i i 4 i
- 1. 2 '
9 i ~
' 1.1 .
l-4 i.
~
' 1. 0 .
-0.6 -0.5 -0.4
-0. 3 -O'. 2 -O'.1 0'. 0
~ n 0'.1 0.2
, ~ 0.3 0.4' 0.5 0.6 AIIAL SEAFE INDEX., Y.1 ' '
- FIGURE 2.2-3 Thermal Margia/ Low Pressure Trip Satpoint. .
STo LDCII - UNIT 1 2d i , Amendment No. p ,4 g Wa 4 e
.O
i l t l i I i PVAR = 2061 . Al a QR1 + 13.85 Try - 8950 QR1 TUNCTION
- 1. 2 i 1.0 '
. (.972,.972)
(.781, .863) N 1 i 0.6< l 0.4<
.0, .235) 1 0.2< j l
0 0 0.2 0.4 0.6 0.8 1.0 1.2 FRACTION OF RATED TRERMAL POWER FIOURE 2.2-4 Thermal Margin / Low Pressure Trip Setpoint Part 2 (Jraction of RATED THERMAL POWER Versus QR1 )
~
i sT. wcII - UwrT 1 2-9 Amendment No. 27,43
_ m._ _ .- . _ .. _ .. . . _ . . . . - _ _ . _ _ _ . . . _ _ _ _ . . . _ . . _ _ _ _ _ ___ - ._ . _ _ . _ _ _ _ . . _ . . _ _ _ . _ . _ _ . _ ._ t TA8tE 3.3-4 5 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRtBENTATION TNIP VALUES c n ALLOWABLE i M FUNCTIONAL UNIT TRIP SETP0 INT VALUES c 1. SAFETY INJECTION (SIAS)
$ a. Manual (Trip Buttons) Not Applicable Not Appilcable
-~
- b. Containment Pressure - High 1 5psig 1 5 psig i
- c. Pressurizer Pressure - Low > 1600 psia > 1600 psia
- 2. CONTAINNENT SPRAY (CSAS)
- a. Manual (Trip Buttons) Not Applicable Not Applicable
- b. Containment Pressure -- Nigh-Nigh 1 10 psig i 10 psig 4
- 3. CONTAlletENT ISOLATION (CIS) *
- a. Manual (Trip Buttons) Not Applicable Not App 11 cable
.L
- b. Containment Pressure - Nigh 1 5 psig 1 5 psig !
- c. Contafrument Radiation - High 1 10 R/hr i 10 R/hi-
- d. SIAS ----- (See FUNCTIONAL UNIT 1 above) ---- ;
- 4. MAIN STEAM LINE ISOLATI0fl (MSIS) t
{ o
.cs.
- a. Manual (Trip Buttons) Not Applicable- Not Applicable g ,' ,.b.. Steam Generator Pressure - Law > 585 psig >585p,fg..
a
= _
'+
- 5. CONTAlletENT SUNP RECIRCULATION (RAS)
,8 a. Manual RAS (Trip Buttons) NotApplicable Not Applicable
$ b. Refueling Water Tank - Low 48 inc'hes above 48 inches above m
tank bottom tank bottom < L _ _ -__ - ___ _ _ - _ _ _ ________ - _ - _ _ _ _ - _ _ _ _ _ _ _ - . . _ - ._. .. . . _ - _ - _ _ _ - _ _ - _ _ _ _ - _ - _ _ _ _ _ .
. ~ . . . . - - _ . , . - .
h TABIE 3.3-4 (Continued) m [NGINLIRLD SAFETY FLAIURL ACTUATION SYST[M INSTRUMENTATION TRIP VALUES
.-e.
,. ALLOWA8lE g IUNCTIONAL UHli_ TRIP VALUE VALUES T.;
, 6. LOSS OF P0utR
' ~
a' . -(1) 4.16 kw Leergency Bus Undervoltage l H (toss of Voltage) . 2900
- 29 volts with a 2900 6-29 volts with a 1 1 5 second time delay 1 1 5 second time delay
- b. 4.16 kw Emergency Bus Undervoltage (Degraded Voltage)
(1) Undervoltage Device #1 3675 6 36 vol ts 'with a. 3675 1 36 volts with a 7 + 1 minute time delay 7
- I minute time delay M l
* (2) Undervoltage Device #2 3592 + 36 volts with a 3592 6 36 volts with a
{ 1812 secund time delay IB 1 2 second time delay
- c. 480 volts Emergency Bus undervaltage {
(Degraded Voltage) 429 + 5-0 volts with a 429
- 5 -0 volts with a 7 1 I second' time delay 71 I second time delay
- 7. AUXILIARY FEEDWATER (AFAS)
. a. Manual (Trip Buttons) Not Applicable Not Applicable
- b. Automatic Actuation Logic Not Applicable Not Applicable E, 'c. SG 1A & IB Level Low 129.01 128.5%
$ 8. AUXILIARY FEEDWATER ISOLATION
- a. Steam Generator aP-High 1275 psid 1281 psid g b. Feedwater Header High AP 1150.0 pgq 1157.5 psid
_. h~ ! n r
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E4 1 4.0 CHEMICAL AND VOLUME CONTROL SYSTEM (CVCS)
- 1. System Name : CVCS F9.3.4 a
. CVCS Pumps
- 2. Pump Identification No.: 1A 1B IC FFg9.3 5
- 3. Pump Type PDP PDP PDP FFg9.3 5 (PDP or Centrifugal)
- 4. Pump Power Supply Bus : 1A 2 18 2 1AB FTbs.3 1
' 440 480 Bus Voltage : (VAC) 480 TTb8.3 1
- 5. a) Pump Design Flow Rate:(spm) 44 44 44 FTb9.3 17 b) Related Discharge Head:(psi) # # # IT 1.2Ad *ttna 3ec yll-
- 6. Maximum Charsmg/ Makeup g/] MJ '
System Flow Rate : (gpm) f (Sum of MAX Charging Flow and Nominal Seal Water injection) i
- 7. Pump shared for High Head No No No F6.3.2.1 (Safety) Injection :
- 8. Pumpe euction cross tied to Spent Fuel Foolf M8 O i
Pump Cooling Water Systems (Include Lube Oil Coolers, Seal Cooling, Motor Cooling, Room Cooling, etc) 9a 1) Cooler Name:
/\/h .
- 2) System "A* Name:
- 3) System *B* Name:
__. =
- 4) Both Systems Raquired: (1 * . n21 redundant)
Ob 1) Cooler Name: l
. l
- ) System *A* Name:
- 3) System *B* Name: , ,
. =. =
- 4) Both Systems Required: (l'e . n21 redundant)
NRC Plant Information Book 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E4-2 l 9c 1) Coolir Ntme:
- 2) System *A* Name:
- 3) System *B* Name:
- 4) Both Systems Required: (1 e . R21 redundant) !
9d 1) Cooler Name:
- 2) System "A* Name:
- 3) System *B* Name:
=
- 4) Both Systems Required: (1 * . R21 redundant) l l
Boric Acid Pumps (Emergency Boration)
- 10. Pump Name : Boric Acid Makeup Pump
- 11. Pump Identification No. 4 1A 1B FFs9.3-5
- 12. Pump' Type : Cent Cent FTb9.3-30 (PDP or p,gnirifugal)
- 13. Pump Power Supply Bus
- MCC 1A-6 MCC 1A-6 FTba.3-1 Bus Voltage : (VAC) 480 440 FTb8.3 1
- 14. a) Pump Design FlowRate:(spm) 142 142 TT" * *n-
[J2#A6
~3 b) Related Discharge Head:(psi) 100 (See 100 (See- dTbH>te-M *Ntmr4). , p ) 7s , f
- 15. Pomp Shutoff Head :(peig) lif centrifugal l t
Boric Acid Supplies
- 16. Tank "A* Name: 1A 6Ae bk FFg9.3-6
- 17. Tank *B" Name : 15 $a m % lt NOTES:
- W~A cont; -^^-J'
. . mye. V alve hatedTnF3 ^..". Tu OO.U NRC Plant Information Book 03/93 .
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E6-1 6.0 INTERMEDIATE HEAD (SAFETY) INJECTION SYSTEM (IHI)
- 1. System Name : High Preneure safety injution System (See Note 1)
General
- 2. Pump (e) crose-tied between Plant None FFs6.3 1 Unite f
- 3. IH1 Pump Success Criteria : 1 of 3 F6.3.2.2.4 (Minimum Number of Pumps) -
IHI Pumps
- 4. Pump Identincation No. 4 1A 1B 1C FFs6.3 1
- 5. Pump Power Supply Bus : 1A.3 1B-3 1AB FTb8.3 1 Bue Voltage : (VAC) 4.16KV 4.16KV 4.16KV FTb8.3 1
- 6. a) Pump D.atian Flow Rate:(spm) 345 (See ses 348 FTb6.3-2 Note 2)
"~~~~~~~ ~~~~"--
- b) Related Dinharge Head:(pei) ------
j .p['.
# sr / .i T,T3 III' b'
- 7. a) Pump Runout FlowRate:(spm) 640 640 640 b FTb6.3-2 b) Related Discharge Head:(pel) 580 580 40
- s. Pump Shutoff Head : (pel) J340'4p ??*'- g- [/78 } 7(( Md Pump Cooling Water Systems (Include Lube Oil Coolers, Seal Cooling, Motor Cooling, Room Cooling, etc)
?
9a 1) Cooler Name: RHI Pump Cooling FM , F9.2.2.2 , 2) System "A" Name: Com nent Cooling
- 3) System *B* Name:
- 4) Both Systeme Required: (i.e.. n21 redundant)
NRC Plant Information Book 03/93
- . . . . . - - . .~ .. . - - - - . . . . . - . . . ..- -~ _ _ . . . . - -
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E6 ,
,! . 9b 1) Cooler Name:
- 2) System 'A' Name:
f
- 3) System *B* Name:
l
- 4) Both Systems Required: (i.e.,31 redundant) ,
4 9c 1) Cooler Name: l
- 2) System "A* Name: I
-- = l
- 3) System *B* Name: .
- 4) Both Systems Required: 0.e., g redundant)
'i 9d 1) CoolerName:
- 2) System 'A* Name: - ~~
I j 3) System "B* Name: . 04.,31 ) { 4) Both Systems Raquired: undant) ' 4 1- NOTES: l- 1. Recire flow is normany provided by at least one HPSI pump. At the discretion of the operator, a portion of the a cooled water from the containment spray system may be directed to the HPSI pump suction. This is not necessary to meet core cooling requiremente but le a preferred ' mode of operation.
- 2. Includes 30 gym bypass flow.
4 1 4 l 1 1 NRC Plant Information Book 03/93
PRESSURIZED WATER REACTOR UNIT. SAINT LUCIE UNIT I E91 l l 1 9.0 LOW HEAD (SAFETY) INJECTION SYSTEM ' l (LHI)/ RESIDUAL HEAT REMOVAL SYSTEM i i l
- 1. System Name : (LH1 function) Low Pressure Safety Injection System
- 2. System Name : (RHR function) Shutdown Cooling System General
- 3. LHI provide euction for HHI N/A.
Pumps during LOCA Racirculation Phase 7 '
- 4. LHI provide euction for IHI No F6.3.2.1.2 Pr.mps during LOCA Recirculation Phase 7
$a. MAX RER Operatina 32s 79.3.s.1 Tempernture: (deg F)
_.~.- 8b. MAX RRR Design Temperature: [h[ .g. b o 4 (des F)
. ... . . .. I iD 6b. MAX RHR Design Pressure: I (psis) 800 ly I a
6c. RHR Suction Line Safety Valve floce}u tt p-o it! y Ni,' D -w l Relief Setpoint:(peig)
- 7. LH1 provide alternate containment spray capability:
a) LHI system have its own spray header i NO /Ib (,
... ...- .~ ._
b) LHI system discharges into Yes FFg6.3 1 the Containment Spray System Spray Hender ?
~ ~ = = - . . - . .
c) LMI system discharge into NO
,f Containment Spray Pump Suction f g 7' {
{
- 8. Pump (s) crose-tied between Plant No FFs6.3 1 Units ?
NRC Plant Information Book 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNI.T I E9 2 9 LHI ( injection function) Pump 1 of 3 FTb6.3 3 Success Critena : (Minimum numtier of pumps) LHI/RHR Pum'ps
- 10. Pump Identification No. 1A IB ' FFs6.3 1 ' -
- 11. Pump Power Supply Bus : 1A 3 1B.3 FTbs.3 1 Bue Voltage : (VAC) 4.16KV 4.16KY FTbs.3 1
- 12. . Pump Shutoff Head : (pei) 17s / FFre.3 3 Pump Cooling Water Systems .
(Include Lube Oil Coolere, seal Cooling, Motor Cooling Room Cooling, etc) 13a 1) Cooler Name: Pump Cooling F9.2.2.2
- 2) System "A' Name: Component Cooling System g F0.2.2.2
- 3) System "B' Name:
4') Both Systems Required: (i.e., g redundant) 13b 1) Cooler Nama:
- 2) System 'A' Name:
- 3) System *B* Name:
- 4) Both Systems Required: (i.e., B21 redundant) .
13c 1) Cooler Name:
=_ .
- 2) System "A' Name:
j
- - =
j
- 3) System "D* Name:
l
- 4) Both Systems Required: (i.e., agi redundant) 13d 1) Cooler Name:
i
- 2) System 'A' Name:
- 3) System *B* Name:
~
- 4) Both Systems Required: (i.e., nel redundant)
Heat Exchanger Secondary Side Cooling Water System
- 14. System 'A' Name : Component Cooling System F9.2.2.2 N -
NRC Plant Information Book 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT I E9-3 16 Syst:m 'B' Name s .
- 16. Both Systems Required : (1.... n21 redundant)
NOTES: b 1 l l J 1 l , l 1 1 e 4 NRC Plant Inforsnation Book 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E101 10.0 SAFETY INJECTION TANKS
- 1. Tank Name Safety injection Taan
- 2. Number of Tanks : 4 FTs6.3 2
- 3. T.S. MIN Water Volume : (uruts) 1090 ft 3 TS3.5.1
- 4. T.S. MIN Nitrogen Pressure: 200 T33.8.1 (peig)
NOTES: i a d i NRC Plant Information Book - 03/93
. PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E11 1 11.0 BORATED ~ WATER SOURCES for HHi,lHl. LHI,- CVCS .
Refueling Water Storage Tank (RWST) ' 1, Tank Nasne : Refueling Water Tank
- 2. Tank identincation No. : 1 FFs6.2 28 3a. T.S. MIN capacity for Operations . 401,800 TSS.I.2.s Mode Esdank : (sal)
=--- _...
9A sb.'I.glal T.S. WIN capacity for operatione : (gal) f-
- 4. Design capacity : (gal) 538,000 / , FTb6.3 2
- 5. 25,000 Unusable Tank Volume (sal) / FTb6.3 2 '
- 6. RWST Supplies CVCS 7 Ya FFs6.2 24
- 7. RWST Suppiles NHI 7 N/A
- 8. RWST Supplies IMI 7 Yes FFs6.3 1 '
- 9. RWST Supplies LHI 7 Yes FFs6.3 1 ,
- 10. Tank crose tied to other Unit'e No FFse.2 28 Tank 7 -
- 11. MIN RWST Level for ECCS Swap 48 la. (See TSTbs.S.4 to Containment Sump 4 (units) Note 1)
(If automatic swapover, use T.S. MIN RWST Level) 4
- Concentrated Boric Acid Storage Tasks (BAST) for Emergency 3 Boration
- 12. Tank Name : Boric Acid Makeup Tank 1
, 13. Tank IdentiGcation No. : 1A IB i FFs9.3 5 i
- 14. MIN No. of Tanks required by 1 of 2 T53.1.2.8 T.S. for Operatione Mode :
15a T.S. MIN capacity for Operatione See Note 2 See Note 2 Mode grLigDh : (gal) 93 Variable TSFs3.1 1 15b I.2131 T.S. MIN capacity for i operatione : (gal)
- 16. Design capacity : (gal) 9700 / 9700 / FTb9.3 18
- . 17. BAST have gravity feed to Yes Yes TS3.1.2.2 injection pump suction 7 i
NRC Plant Information Book i 03/93
-+-, - r- , - ,,, . - , , , . ~ . . ~ - - - .
- GI 3 '
S".. Ci ' k N BAk" V0..E nm acama l
/ll,f fI \Y
/ /
} h[))////
Y ' ACb P ABLE j, l/
///l
/d '
il" g O/v/
/
/g W -
>e /,/, /
1 01250)-
- n. g s /
V / / / I UNACCE PTABLE V 3 oPERATl oN N/[
, b */ .
l(8350) 8 ,l U U U U U U. U (4196 PPM) (4546 PPM) (4895 PPM) (5245 PPM) (5595 PPM) (5944 PPM) (6294 PPM) SMED IM Colt (vtIbuic add) ST. 1.UCIE - UNIT 1 3/4 1 17 Amendment No. 27,94 MO 0M YY ANf8'1
PRESSURIZED WATER. REACTOR UNIT-SAINT LUCIE UNIT 1 Ell-2 D '
- 18. Oth:r T.S. Ssurces far Emerg:ncy 1) .M Boration (other than BAST or RWST) : 2)
NOTES:
- 1. Above tank bottom.
2. Tank 1A or 1B in range of $595 ppm to 6119 ppm boron (Source TSFgs.t.1). a J t a. d I i i i l 4 i. 1 i i I a d NRC Plant Information Book . 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E 12-1 12.0 CONDENSATE, FEEDWATER, and CIRCULATING WATER SYSTEMS Condensate Pumps
- 1. Pump Identiacation No.
- 1A IB IC F10.4.6 &
FFs10.1 2a
- 2. Pump Power Supply Bus : 1A 2 1B 2 FTba.3 1 Bue Voltage :(VAC) 4.16KV 4.16KV g/ FTb8.3 1
- 3. Motor shared with Condeneste N/A Booster Pump 7
- 4. Pump Capacity : (% Full Power per Pump) g e4
- 5. Minimum No. Pumpe required for single Main Feedwater Pump /
operation :
, ( 6 Y ,.
- 6. Pump Shutoff Head : (pel) g]
Condensate Booster Pumps 1 i
- 7. Pump Identification No. N/A
- 8. Pump Power Supply Bus :
B.us Voltage : (VAC)
' 9. Pump Capacity : (% Full Power per Pump)
- 10. Minimum No. Pumps required for single Main Feedwater Pump operation :
- 11. Pump Shutoff Head :(pei)
Main Feedwater Pumps
- 12. Pump identincation No. t 1A 1B FFs10.1 26
- 13. Pump Power Source : (Motor or electric electric F10.4.6 .
Steam) l
- 14. Pump Electric Power Supply Bus: 1A1 1B1 FTb8.3 1 Bus Voltage : (VAC) 6.9KV 6.9KV
- 15. Pump Capacity : (% Full Power e per Pump)
NRC Plant Information Book 03/93
PCESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT I E12 2 Other Feedwater' Injection Pumps ,
- 16. System Name : N/A
- 17. Pump IdentiAcation No. .
1s. Pump Power Source :(Electric or Steam)
- 19. Pump Electne Power Supply Bue:
Bue Voltase :(VAC)
- 20. a) Pump h Flow Rate:(spm) b) Related Discharee Head:(pei) '
Condenser Circulatlas Water System ! 1
- 21. Etimate usat sink : Atlantic Ocean FFs9.2.7.2 NOTES:
i l l l i 1 NRC Plant Information Book 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E13 1 13.0 AUXILIARY FEEDWATER WATER SUPPLY SOURCES 1.0 Water Source #1
- n. Source Name: Condensate Storage Tank Unit 1 F Fg10.1 2c &
10.5 2
- b. Source hard-eined to AFW Yes FFg10.5 2 Pumpe Suction 7
- c. MIN T.S. Volume : (units) k
- d. MIN T.S. Volume : (gal) 116,000 TS3.7.1.3
- e. Design Capacity : (gal) 250.00 FTb9.2 16 ,
- f. Tank refillable by : (Name) 1) b neralise Water System M FFg10.1 2c
- 2) Condeneste System Unit 1 FFg10.1 2c
- 3) P N ^ ,# 'Un F Pg1
=
- 4) D ystem Unit 2 la
~
1 2.0 Water Source #2
- n. Source Name: Condensate Storage Tank Unit 2
- b. Source hard-eioed to AFW Yee Pumpe Suction 7
- c. MIN T.S. Volume : (units) h
- d. MIN T.S. Volume : (gal) 307,000 Unit 2 TS3.7.1.3
- e. Design Capacity : (gal) 400,000 Unit 2 FTb6.9.2 A- ,
- f. Tank refillable by : (Name) 1) he neraliser Water System h Unit 2 FFg10.1-
- 2) Condensate System Unit 2
,. Unit 2 FFr10.1
- 3) e e og S e 2a nit 1 Ur FF p I
f
/ : F j c NRC Plant Information Book 03/93
, PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT I E13 2 3.0 Water S:hres 03 I
a.. Source Name:
- b. Source hard-otoed to AFW Pumpe Suetion 7
~
- c. MIN T.S. Volume : (units)
- d. MIN T.S. Volume : (gal)
- e. Design Capacity (gal)
- f. Tank refillable by : (Name) 1) 2)
3) 4) 4.0 Water Source #4
- n. . Source Name: '
- b. Source hard-mined to AFW Pumps Suction 7
- c. MIN T.S. Volume : (units)
- d. MIN T.S. Volume : (sal)
- e. o ign capacity :(sal) l
- f. Tank refillable by : (Name) 1) 2)
s) 4) ~
\
1 NRC Plant Information Book 03/93 ! I
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E13 3 5.0 Wat2r Sizrce c5
- a. Source Name:
- b. Source hard eined to AFW Pumpe Suction 7 j 1
- c. MIN T.S. Volume : (unite)
- d. MIN T.S. Volume : (gal)
- e. Design Capaci:y : (gal)
- f. Tank refillable by : (Name) 1) 2)
i
- 3) )
l
- 4) ._
1 l 6.0 Water Source #6 ' l
- a. Source Name:
- b. Source hard-paced to AFW Pumpe Suction f
- c. MIN T.S. Volume : (unite)
- d. MIN T.S. Volume : (gal)
- e. Design capacity : (gal)
- f. Tank refillable by : (Name) 1) 2)
3) 4) NOTES: i NRC Plant Information Book 03/93 -
.4am.p d _ m m dH-4+ d-dE ---M* J PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 'E14-1 '
14.0 AUXILIARY FEEDWATER SYSTEM 1
- 1. . System Name : Auxiliary Feedwster System l System Configuration i j
- 2. Pump IdentiScation No. : 1A 1B 1C FFg10.1 2c l
& Tg10.5 2
- 3. Steam Generator ID for all 1A,1B 1A,1B 1A,1B FFt10.1 2c possible injection lineupe :
- 4. Pump Auto initiate f Yes Yes Yes F10is.2
- 5. Pump Crees-tied between Plant No, No No FFs10.1 2s Unite :
l Auxiliary Feedwater Pumps
- 6. Pump IdentiScation No. : 1A IB IC FFs10.1-2c 7 Pump Type : M M S FFs10.5 2 l (jdetor/14eam/Riesel) l (en t r e d.
- 8. Pump Electric Power Supply Bus: 1A-3 18 3 FTb8.3 1 pg g Bue Voltage :(VAC) 4.16KV 4.16KV g 0C FTbs.S.1 )
- 9. Steam Generator ID suontrinz 1A,15 FFs10.1-la 11aam to Turbine
? l
- 10. MIN Design Steam Pressure for 50 F10.5.2 Turbine Operation : (peig)
I
- 11. a) Pump Regig Flow Rate:(spm) 225 / 338 / 600 / FTb10.1-1 i b) Related Dischar e poi) 1185 [ :185
[ 1185 / - q 12. Indicate all sources of water #1,#2 #1,#2 #1,#2 FFs10.1 2e sTi- r supply to each pump suction : (Water Source Number from Sect. 12) 1
- Pump Cooling Water Systems (Include Lube Oil Coolers. Seal Cooling, Motor Cooling, Room Cooling, etc) i .
NRC Plant Information Book . 03/93
PRESSUkIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E 14-2 131 1) Coohr Name: gg g
- 2) System "A" Name:
- 3) System "B" Name:
- 4) Both Srstems Required: (i.e., g redundant) 13b 1) Cooler Name:
- 2) System 'A' Name:
- 3) System *B' Name:
- 4) Both Systems Required: (i.e., g redundant) 13c 1) Cooler Name:
- 2) System 'A' Name:
< 3) System *B* Name:
- 4) Both Systems Required: (i.e., g redundant) 7 13d 1) Cooler Name: -
-_ _ _ . ~ . _ . -
- 2) System "A* Name:
- 3) System *B' Name:
- 4) Both Systems Required: (i.e., a redundant)
NOTES: i 1 J w 4 NRC Plant Information Book 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 El51 1
'15.0 MAIN STEAM SYSTEM i
l
- 1. Number of Steam Generators : 2 FFs10.1 la
- 2. Steam Generator model :
gf, p g,y /h d [e a k /* f'
- 3. Total Number of Main Steam 2 FFg10.1.la Lines:
- 4. Total Number of Main Steam 2 FFglo.l.la Isolation Valves s
- 5. Minimum T.S. Main Steam Code 1000 TSTb4.7 1 Safety Valve Setpoint : (peig)
- 6. Steam Generator Blowdown Yes F10.4.7 System available 7
- 7. Turbine Bypase Valves Total 45 % F10.4.4 Capacity : (% Full Steam Load)
- 8. Atmospheric Dump Valves Total e Capacity (% Full Steam Load) l NOTES:
l l i I l i I NRC Plant Information Book 03/93 -
g PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E 16-1 I 16/J REACTOR CONTAINMENT i A
?
- 1. Containment Type : Larse Dry
. 2.' MIN T.S. Containment Free 2.5x10 6 TSE.2.1.1 I
Volume : (cu ft) - l ',2 Total Zirconium Inventory in 44.700 FTb4.2 1
]- Core : (Ibe)
{ 4. Design Temperature Limit : (OF) 264 T55.2.2
- 5. ' Design Internal Pressure Limit: 44 TSE.2.2 '
(pois) 6a T.S. MAX Leak Rate : 0.50% TSS.6.1.2 (% y.algma or 1.Esight per 24
- a hre) 6b T.S. MAX Leak Rate Pressure: 39.6 T53.e.1.2 (pois) s i
i Hydrogen Recombiners 1 l
- 7. Number of F - -- " - . :
2 F6.2.5.2.1 1
- 8. Recombiners installed ? YES F6.2.5.2.1 l 9. Containment have Ignitors ? NO 6.2.5.1
- 10. Hrdrogen Purse System installed? YES F6.2.5.2.2
- 11. Hydrogen Purse filtered ? YES F6.2.5.2.2 .
i i i NOTES: J t i 1 d i 4 I i , i
.NRC Plant Information Book 03/93 i
. , , , _ - - , ,-- - - , ,+n- ,, -
PRESSURIZED WATER REACTOO UNIT-SAINT LUCIE UNIT I E17 1
. 17.0 REACTOR CONTAINMENT COOLING SYSTEM l
- 1. System Name : Containment Cooling System F6.2.2.2.2 System Configuration
- 2. Total Number of Cooling Unite : 4 F6.2.2.1
- 2. Containment Cooling System used Yes F 8.2.2.1 for Containment Depressurisation during LOCA or DBA accident conditions f IM2 - Do ngg complete this section.
Success Criteria (Containment l Depressurization)
- 4. Minimum Numbs of Fans per Mf fleg 43) l Cooling Unit:
l S. Minimum Number of Cooler Units with No Containment Sorays : 4 of 4
/ F6.2.2.1
- 6. Minimum Number of Cooler Unita with 1 Containment Sorav Train 2 of 6 / F6.2.2.1 Ooerattnr :
Cooler Units UNIT-A , UNIT-B UNIT-C UNFF D , l 7a. Unit Identification No. . 1A 18 1C 1D FFs9.4 1 8b Fan Electric Power Supply Bus: 1A2 1A2 1B2 1B2 FTb8.3-1 l 480 440 l Bus Voltage : (VAC) 480 4so FTbs.3-1 ' UNIT-E UNIT-F UNIT-G UNIT-H 7b. Unit Identification No. :
% [M= gg [M Bb. Fan Electric Power Supply Bus:
. . - . . _ . -~ _ _ .- _. .. ..
Bus Voltase : (VAC) l 1 NRC Plant Information Book 03/93
PRESSURIZED WATER REACTOR UNIT. SAINT LUCIE UNIT 1 E17 2 Cecist Uzits Cectlzg Wetar Systsm
- 9. a) System 'A" Name:
Component Coolinspystem. [#kb F9.2.2.2 b) System *B* Name:
. . ~ .. -
c) Both Systems Required: (i.e., agi redundant) Noizs: f, & MN 1, fk k & m Use c*iu yy c c W. a NRC Plant Information Book 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UN,IT 1 E18 1 i i 18.0 CONTAINMENT SPRAY SYSTEM
- 1. System Name : Containment Spray System System Configuration
- 2. Containment Spray designed for Yes F6.2.2.1 containment depreneurisation f 4
- 3. RHR Heat F.xchangere shared Yes F6.2.2.2.1 with Containment Spray ?
- 4. Can RHR system provide alternate containtnent spray capability:
. . - = . - . . . *
- a. LHI/RHR system have its No FFs6.2 2s own spray he'ader?
9E
- b. LHI/RHR erstem discharges FFs6.2 28 Into the Contalgrnant Spray ,
System Spray Headerf 9.E
- c. LHI/RHR system discharge No FFs6.2 28 into Containment Spray Pump suction?
- 5. Pump (s) Croes. tie between Plant No FFs6.2 28 1 Unite ? )
l 4 Success Criteria (Containment j depressurization) 4
- 6. Minimum Number of Pumps with
& Containment Coolers :
[
- f. Recirculation and Quench Sorav ,
Confiruration ONLY: y , Minimum Number of Pumps with v
& Containment Coolers :
NRC Plant Information Book 03/93
PRESSURIZED WATE3 REACTOR UNIT-SAINT LUCIE UNIT 1 E18 2 Ccatainment Spray Pumps I PUMP 1 PUMP-B PUMP-C PUMP-D an. Pump Identification No, s 1A IB h[ h FF 6.27s ca. Pump Power Supply Bus: 1A3 IB3 FTb8.3-1
- - - - . ~ . .- -- -. ..
4 Bus Voltage : (VAC) 4.16KV 4.16KV FTb8.3-1 11a Pumps used for containment l Yes Yes F6.2.2.2.1 l spray via Sump recarculation? l 12s Pumps Located IN1!DE No No l FFse.2 28 l rentainment: Ptre4P-E PUMP-P PUMP-G PUMP-H 8b, Pump Identification No. : gp gp [p gp 9h. Pump Power Supply Bus: 1 Bue Voltage : (VAC) l R 11b Pumps used for contalamant spray via Sump recirculation? 12b Pumpe Located [Eg&E Containment: f Pump Cooling Water Systems (Include Lube Oil Coolers. Seal Cooling, Motor Cooling, Room Cooling, etc) 13a 1) Cooler Names Containment Spray Pumps f(ddethC" .*
- 2) System "A" Name: Component Cooling Systeau F9.2.2.2
- 3) System 'B* Name: W MS
- 4) Both Systems Required: (I e . n21 redundant) 13b 1) Cooler Name:
- 2) System *A" Name: l
- 3) System 'B' Name:
- 4) Both Systems Required: (3 e . n21 redundant) 13c 1) Cooler Name:
- 2) System 'A" Name:
- 3) System 'B' Nami:
- 4) Both Systems Required: (i s., agi redundant)
NRC Plant Information Book . 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E18-3 13d 1) Cooler Name: '
- 2) System "A" Naw.e:
- 3) System "B" Name:
- 4) Both systems Required: (1 * . 221 redundant)
Heat Exchanger Secondary Side Cooling Water System
- 14. System "A" Name :
Component q$ysteen, , FTg9.2 2 m_ f
- 15. Bretem 'B' Name : - -
- 16. Both Systems Required : (i.e., gal redundant)
NOTES: 4 l I l 1 1 t
- NRC Plant Information Book -
03/93
PRESSURIZED WATER REACTOR UNIT. SAINT LUCIE UNIT 1 E19 1 19.0 ALTERNATE AC POWER SOURCES MI Class IE Emergency Diesel Generator (EDG)
- 1. Total Number of EDG On. site : FFgs.3.la 4 (See Note 1)
- 2. Number of EDG for this Plant 2 (See Note Fs.3.1.1.7 Unit i 1)
- 3. Number of EDO shared between 98.3.1.1.7 this Plant Unit and other Unitet g DG.A DG.B DG.C DG.D 4a. EDG Identification No. : 1A IB [k FFs8.3.la Sa. Diesel Manufacturer (s) : , , [h DG.E DG.F DG.G DC.H 4b. EDG Identi6 cation No. .
h E[M IUYM N M M 's 5b. Diesel Manufacturer (s) :
- 6. No. of Dioeel Fuel Storage Tanks : 2 F9.5.4.2 1
- 7. Capacity of Storage Tanks : (gal) 88,350 F9.5.4.2 1 (Total) ,
I l Class 1.E Diesel Cooling Water Systems
~
en 1) Cooler Name: SettCosaed 14 7 -[Mrrj F9.5.5.2, 9.5.5.3
- 2) System "A" Name:
- 3) System *B* Name: '
l 4) Both Systems Required; (i.e.. En redundant) ab 1) Cooler Name:
- 2) System *A* Narne:
- =
- 3) System *B* Name:
- 4) Both Systems Required: (i.e., ng redundant) l 8c 1) Cooler Name: '
- 2) System *A' Name: -
- 3) System "B" Name:
- 4) Both Systems Required: (i.e.. R21 redundant) l NRC Plant Information Book 03/93 i
)
i
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT I E19 2 8d 1) Cooler Name:
- 2) System "A" Name:
- 3) System "B" Name:
- 4) Both Systems Required: (1 e . Del reduMant)
Other Class IE Emergency Power Sources
- 9. Source Name :
- 10. Source IdentiScation No.
- 11. Electrical Bus ID Ties - All Possible Direct Tlas Other Class IE Emergency Power Sources Coollag Water Systems 12a 1) Cooler Name:
- 2) System *A* Name:
- 3) System *B* Name:
- 4) Both Systems Required: (I e-. G21 redundant) 12b 1) Cooler Name:
- 2) System "A" Name:
- 3) System *B' Name
- 4) Both Systems Required: (I * . n21 redundant) 12c 1) Cooler Name:
I 2) System "A" Name:
- 3) System *B' Name:
- 4) Both Systeme Required: (1 e . E21 redundant) 12d 1) Cooler Name:
- 2) System *A* Name:
- 3) System *B" Name:
- 4) Both Systeme Required: (1 e . Del redundant)
NRC Plant Information Book 03/93
E19 3 PRESSURIZED WATER REACTO3 UNIT-SAINT LUCIE UNIT I . Nea-S:fety Gr:de Alters:t2 Pawsr S:ppilts
- 13. Source Name,:
- 14. Source Identification No. :
- 15. Electrical Bus ID Ties - All Possible Direct Ties :
- 16. How Interconnected :
5" .- NOTES: YA
- 1. Two engines in tandean, with senerator in the nuddle. Yaivelisted is # of generators. Then are a total of 4 dieself for unit 1 (Source F9.5.5.2).
r . l
- ~ - .
he [ht Ydih j d A 4lhe V Tf4N Cn h Crn t fo t4 gg o*5
$tc(SJ4Yy by caw ecfy ray g,aiyy jy pp '
W f, Ie s4urrof D c re Wr4 e l$r01 ac ess fps-r cpily #e r
# H.tr.
l,/>f 4sr & Jc p I.+I'! W #) , (Qu.(CClFT,3e i 03/93 l NRC Plant Information Book
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E20-1 20.0 COOLING WATER SYSTEM
- 1. System Name : Intake Cooling Water system System Configuration
- 2. System Purpose : Provide a heat sink for the component cooling system, F9.2.1.1 turbine cooling, and steam generator blowdown cooling system, etc. gj y
- 3. Systern shared between plant Wes h eape/ 4 -
) FFr9.21 b' Units 7 g
, 4. System Safety Related 7 Yes F9.2.1.1
- 6. Closed or Open System 7 Open F9.2.1.1 Success Criteria - Each Plant Unit (Accident Conditions)
- 6. Minimum Number of Pumps : 1 of 3 F9.2.1.3.2
- 7. Minimum Number of Heat ' & q' g J,2 Exchanger :
)$fA ~
Pumps PUMP A PUMP B PUMP-C PUMP D a n. Pump identification No. r 1A IB IC FFg9.2 1 9a. Pump Power Supply Bus: 1A3 IB3 1AB FTbs.3 1 Bus Voltage :(VAC) 4.16KV 4.16KV 4.16KV TTb8.3-1 PUMP E PUMP-F PUMP G. PUMP-H 8b. Pump Identification No. - kk [k g 9b. Pump Power Supply Bus: Bus Vol. age : (VAC) NRC Plant Information Book 03/93
. . . _ _ . _ . _ - . . . ~ . . . . - . _ _ _ . _ , . - . _ _ . _ _ _ - _.
1
- PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E20-2 Pump Cecilts WCter Systems ,
(include Lube Oil Coolere, Seal cooling, Motor CoogRoom Cooling, etc) 4 ' los 1) Cooler Name: n'ZW) - - . A g 1 4 g j ggp;g
/
F9.2.1.2 '
\ - _=== . . . . . . _ . _
- 2) System *A* Name: p f hh I Vs: '
h ' j - ~ y rp r.-y *- 4 3) System *B* Name:
- 4) Both Systems Required: (i.e., g redundant) l 4
10b i) Cooler Name:
\
- 2) System "A" Name: ,
4
- 3) System *B' Name:
j
- 4) Both Systems Required: (IA, g Mundant) 10c 1) Cooler Name:
- 2) System *A* Name:
- 3) System "B* Name- I
~
(i.e, a redundant)
- I
- 4) Both Systems Required: i I
1 ' lod 1) Cooler Name:
. 2) System "A" Name:
- 3) System *B" Name:
- 4) Both Systems Required: II'*" g redundant)
]
. l 1 . liest Exchanger Secondary Side Cooling Water System I i
- 11. System "A" Name :
i l 12. System "B" Name : , 13. Both Systems Required : (i.e., g redundant) 4 4 4 NRC Plant Information Book 03/93 -
. PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT I E20-3 WARNING (Generic Statement) l This matrix below list all cooling water system pumps and associated heat exchangers which can be used to maintain a specific component operable usine the full esoability 1
of system cross-ties and interconnections. i i l It should be recognized that under normal circumstances, the cooling water system may ! be operated as isolated " trains" to prevent common mode failures. Consequently, cross tied alignments of the system may not be procedurally permitted. !
- 14. Heat Load Matrix tD Reference Matrix Pump IdentiBcation Matrix Heat Exchanger ID No ID Identiacation No A 1A (Soune FFs9.s 1) A B IB (Source FFs9.s 1)- B .
C 1C (Source FFg 9.s 1) C D D z a F F G G H H
- 15. Heat Load Matrix PUMPS HEAT EXCHANGER COMPONENT NAME & ID A B C D E F G H A B C D E F G H Component Cooling Heat Exchanger X X X 1A (Source FFg9.s.s)
Component Coolir.g Heat Exchanger X X X IB
.- I NRC Plant Information Book 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E20o4 4
~
J W 1 i t 1 i I i - l. . i i & I 4 4 , 1 , n
- i i
4 i i s 4 i i - l 1 i-I i- . NOTES: I
, I NRC Plant Information Book-03/93.
. PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 E21 1 !
21.0 COOLING WATER SYSTEM
- 1. System Name :
Ccmponent Coolingpystem ,, System Configuration 1 j
- 2. System Purpose : l Provide heat eink for auxiliary erstems, provide an F9.2.2.1 -
1 intermediate barrier between the reactor coolant and the intake cooling water systems, provide heat sink for safe'ty related components associated with reactor decay heat removal. 1
- 3. System shared between plant f - '-*" - " b R. ,, FFs9.2.2 Units ?
)Y[
- 4. System Safety Related ? Yes F9.2.2.2 1 l
5 Closed or Open System 7 Closed F9.2.2.2 Success Criteria - Each Plant Unit (Accident Conditions)
- 6. Minimum Number of Pumps : 1 of 3 F9.2.212
- 7. Minimum Number of Heat 1 of 2 F9.2.2.3.2 Exchanger : i Pumps 9
PUMP.A PUMP.B PUMP.C PUMP.D sa. Pumr Identification No. lA IB 1C FFg9.2 2 l l 9s. Pump Power Supply Bus: 1A3 IB2 1AB FTbs.3 1 Bus Voltage :(VAC) 4.16KV 4.16KV 4.16KV FTba.3 1 PUMP.E PUMP.T PUMP.G PUMP.H ab. Pump Identification No. h M 9b. Pump Power Supply Bus: Bus Voltage : (VAC) NRC Plant Information Book 03/93 -
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT I E21-2 Pump Cecil:g Wctar Systems (include Lube Oil Coolers, Seal cooling, Motor Cooling, Room Cooling, etc) los 1) Cooler Name:
- 2) System *A* Name:
- 3) System "B' Name:
- 4) Both Systems Required: II'**' agi redundant) 10b 1) Cooler Name:
- 2) System *A* Name:
. = 1
- 3) System "B" Name:
- 4) Both Systems Required: (I e . Dal redundant) 10c 1) Cooler Name:
- 2) System "A" Name: J
- 3) System "B" Name:
1
- 4) Both Systems Required: (1 e . D21 redundant) 10d I) Cooler Name:
- 2) System *A* Name- '
\
- 3) System *B* Name:
- 4) Both Systems Required: (1 e . n21 redundant)
Heat Exchanger Secondary Side Cooling Water System '
- 11. System *A" Name : Intake Cooling Water System F9.2.1.2
- 12. System "B* Name :
- 13. Both Systems Required : (i.e . n21 redundant) 1 4
?
NRC Plant Information Book 03/93
1 PRESSURIZED WATER REACTOR INT-SADM LUCIE L' NIT 1 E21-3 ) WARNING 9 (Generic Statement) This matrix below list all cooling water system pumps ud associated heat *wehengers which can be used to maintain a specifie component operable neia- the full esa=Mihv of sv=ta= cross-ties and . It should be recosmzed that under normal circumstances, the cooling water system may be operate isolated
- trains
- to prevent common mode failures, i
Consequently, cross-tied ahamments of the system may not be procedurally permitted. I l l 1 l l l
- 14. Hans Land Manix ID Redesumso Meeris Puesp h
- No Meeria Hans Bashauger ID ID Idenadsmanse No A IA A 1A a la a la C IC I C I D D E E F F G , G 1 H E l i
i
- 15. Heat Load Metris '
~ I i PUMPS HEAT EXCHANGER l COMPONDrr NAME & ID A B C D E F G H ' ' A B C D E F G H shuudown Heat t ' . lA
- X X I X X
, X ) i CH- Fan Cooler I A
- X X i X X X
Na- Fan Cooler 1B
- X X X X X
tow Pressure Sefery Impesmoe Pusup i A
- X X 4
X X X , High Pressure Sefury lavesmos .'., I A
- X X X X
', X NRC Plant Information Book 03/93
l PREhSURIZED WATER REACTOR UNIT-SALNT LUCIE UNIT 1 E21-4 i Comaammees Sprey Pump 1 A
- X X X X 5
X 1 Shadown Heat reaanper IB X
- X X '
X X Conseiamses Fan Coo 6er iC X
- X X X X
C- ----- - Fem Cooler ID X
- X X X X
, Low Pneous safery lesees.oa Pump IB X e X x x , X 4
High Prenews Safesy insecoon Pump IB X
- X X X I
' 1
'ha- Spesy Pusip IB X
- X X X '
X -- n High Pnemass Andegr tapeamos Pump IC m y_ w ._ :w .,. X
*- X,
.y u . w_ . 1.. u.._ -
X X 3
~
f 4 g i RCP Pump (seel R&) 1 A1, I A2, X X X X X j 131.IB2 1 RCP Motor I AI, I A2.151, IB2 X X. X X X i l I Fuel Pool Hess 8- ' ;-- X X X X X l l Seese Osamenor Blowdows Sempting X X X X X Cosemi Rod Drive Air (** X X X X X Quemeh Tsak Heat Treaefer Unis X X X X X l ' 1 Borie Acid C - . I A. IB X X X X X l Wome Coneessnesor X X X X X Wome Oes Courpressors I A. IB X X X X X Landown Heat r.a- ... X X X X X i 4 NRC Plant Infoetion 3cok 03/93
l PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 1 l E21-5 l h NOTES: An
- denotes the preferred path p
1. HPSI pump 1C le supplied by header B by adadaistree6ve poseedures. j/ i l
)
P NRC Plant Infor: nation Book 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCI UNITI EA 1 Appendix A Abbreviations c . i AFW Auxiliary Feedwater CVCS Chemical and Volume Control System i i DBA Design Basis Accident des Degree (Temperature) ECCS ' Emergency Core Cooling System j-EDG Emergency Diesel Generator i ESF Emergency Safeguards Features spm Gallons per Minute HHI High Head (Safety) Injection i IHI Intermediate '.ie.ud (Safety) Injection LOCA Loss of Coolant Accident i . LHI Low Head (Safety) Injection MAX Maximum MIN . Minimum " \
' NDTT Nil Ductility, Transition Temperature NRC Nuclear Regulation Commission NSSS Nuclear Steam Supply System i
PDP Positive Displacement Pump PORY Pneumatic Operated Relief Valve i psi Pounds per Square Inch PWR Pressurized Water Reactor i PZR ?rt.ssurizer RCP Reactor Coolant Pump RCS Reactor Coolant System q RHR Residual Heat Removal RWST Refueling Water Storage Tank sq in Square Inch VAC Volts - Alternating Current t I i Reference Notations F Final Safety Analysis Report 4 i (FSAR) FFg FSAR Figure
- FTb FSAR Table TS Technical Specifications (TS) -
TSFg TS Figure '. TSTb TS Table TSBasis TS Basis Section D Licensee's Simplified Drawings (Provided in Section D of Plant
- Information Book) ;
i. 9 i. NRC Plant Information Book ) t , 03/93
- t 6
l I . i U.S. NUCLEAR REGULATORY COMMISSION ! yg,AR Regu
+ o
\ A a % ! O I T C p s,% /. l l
+++++
, PLANT INFORMATION l BOOK l l SAINT LUCIE 2 i l VOLUME I- EMERGENCY RESPONSE l -
1 e 4 SAINT LUCIE UNIT 2 i
. PLANT INFORMATION BOOK i
5 PREPARED BYr University ' of Maryland Student Researchers EDIYED BYr John MacKinnon, AROD msnmmCY RESPOMBE SECTION BYr , NRR a ' DRANINGS PROVIDED BYr REVIENED BY l
. ASSEMBLED BY: .
Joe Sebrosky, AROD REVISION O
1 l j NRC PLANT INFORMATION BOOK SAINT LUCIE UNIT 2 6 1 TABLE of CONTENTS A. Facility Statistics B. Emergency Response information I C. Plant Description Summary D. Simplified Plant System Diagrams E. Detailed Plant Systems Data d i 4 9
i r 1913/93 . A. FACIIJTY !rrATIN!1CS GENERAL INFORMATION Ples Name: Seimt imde Unk 2 Donket No.: 30 309 owaarlopweser- Flodde Power & IJght Co. , 1 Other Pleans On sies: Salme Imda Unit 1 1
, Sies Losemon/Addenes 12 amass mesehenst of Fort Pierce, Flodde Pasility Licensed Thenmal Power Limit: (Ref: Weemme). 2700 MWt a
Main Tusbane Generseer Resed Ousput: SDSMWe W) Ubiasse Heat Sink Supply to Main Turbine Condammer- Adamele Oceam i Ubiennes Heat sink . Supply to ECCS Semse Water Syeesse: Aslemele Ocean i FWR Type: 24eep
- Naas Vendor f*h Ememeerdmg
)'
- 1bsbane Gemarator* Wendughouse Assh. Eage.
- Eheses Services. Inc.
1 Consenseent: Floride Power & IJdit, Eheseo Service i Does of Opesanag Usease issuanne 96/10/E5 j Does of Caeumarosal Operemos. 98d8843 4
- F1. ANT BUMMARY l
Comenismasm Type: 1arge Dry Design Pouemur 44 pds j i sinam osmarmaar ' Sesam Gemmenear Model: Number of Sesam Generneore Number of Mais seensa leelsence Valves: 2 f.Il01 No.of Fuel Assembises 217 " I No. of Fuel Rode per Assembly: 236 s i I Seist Lassie Unit 2. Rev 0
IGY13&3 Chaminal and Valumns Casand Synssa No. of CasanAspel Puespe: . Nome
- No. of Posisive E,' : Pumps: 3 e\
Pu , % now R e: *:a w, v 2J/o -27 g/LJ Hash Hand (Safstri lanamnes Symasm No. of Puespe: Name
- Punip Design Flow Race
- N/A 4
Pump abutorHead: N/A s i 1 i kaarmadissa Haqd faginav) Iniaansa Svanam o No. of Pumps:
,u., % n.w Ra.e:
Pump shusoR Head: 2
. 4/f-w$ /
> 13N pel i _.
Imw Hand fSatavi Ippmanna Symann flJE 1 No. of Pumpe: 2 . j 1
. Pump shunotHead:
i Ranidual Heat Ramaval avamm
. RNE and LMI utillmas she sans W7 Yes
- No. of Pumpe
- 2 1 Pump ah===KHead:
'1h, ,
- P ws,.,
ra ! No. of Punys: Puey ShanotHead: iz.- 3 g%..;f $r. 73 ] c h assaemr 700 i No.of Pusspe: Name Pump ShusetHead: N/A , Main Feedwasar i p No. of Pumpe: 2 hamane Dr6,em Startae Pandwater No. of Pumps: Name t Ana Feedwater Synname 4
, Numiber of Maeor Driven Puespe: 2 Capeony:
300 gun , . Neuber of Twbies Driven Pusups: 1 j capaeny: sTe esa - Aux Fandwater Aussiv Souna - Nees: Camdemance Storage Task Unit 1-250,000 gals Nesse: -
- Camdsammee Seeruse Tamk Unit 3 'an, man gals
..l 4
1 0 . 4 2 Saint 1meie Unk 2. Rev 0
10/13/93 A. FFrf ETY frATIft1CS l
} l Seist Imde Unk 2 5 309 . - y- ~ y . ,.y
.' 6 n . ,f . ,k _ .,}
Flodde Power & IJoht Co. Sebd Imcie Unk i 12 edius neuesent of Fort Pierce, Flodde Oter:Ijeense) 27e9 MWt i 350 MWe (gress)
) [ [fi Servise C'eser sysemen: Adensic Oceae 24eep cer% i y y Iheses Serdcas. loc. y 7 7 u h .\ s ;
Findde Power a % h h , 1
)
estem 1 f acess
-7 ..
.- .o
- 7.. _.l... _ _a .__ .
Imros Der i i 44 pds i f , 3) 1"' A D .va: 2 82 - ., tl -_.- .,d._.11_
. __ __ { j 3g;-~
236 - e 1 i l !
.- ,. ..L, ,k .R, , _P l
* \
l f seiet tacie Unit 2, Rev 0 ty u . I
- h. Ird. 2 - ;
I
f - -- ~, , e 4 s,
/ i PLANT INFORMATION BOOK .
~ i.* p# SA!NT LUCIE 2
'/
VOLUME I- EMERGENCY RESPONSE , i 1 . I i 4- - V V~ ~~) e g e er ? , E
, . . . , . ..-.-....,,m, ,m-,_-,,.. - , _ , ,.,v--,,,-.---.-.,., .- . -,w.m.~. .--,,,..,-.4.,,,,,,,,,,,_..,,.,% . . - . , , , , . . . , , , . , , . , _,,.,,.,.,.__.,.,,g , , . . , , ,,.y , . . , , -., - - - 3
- . . ~ .
unira th/4OhW l B. EMERGENCY RESPONSE INFORMATION 1 .
- SAINT LUCIE UNITS 1 and 2 l
! TABLE of CONTENTS 4 . l } Emergency Response Facilities i l Site and Population Emergency Response Officials l l Appendix: Drawings, Charts and Maps ' 1 1 l . i . i I t i i d i l NRC Plant Information Book ,
1 i i > B. EMFRGENCY RFRPONSE INFORMATION-Plant Name: St. Lucie Unit 1 & 2 n . EMERGENCY RESPONSE FACH211ES
, TSC: Technical Support Center is located adjacent to the Unit 1 Control Room at the St. lacie Plant. Reference Attachment 1. ,
j 1
- EOF
- . Emergency Operations Facility is located at the Midway Substation l approximately 10.1 miles due west of the power plant at the intersection of state Road 712 and I-95.
i (Refe cace: 10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) p . ! "OSC: Operational Support Center is located in the North Service Building in the Conference Room on the second floor. '
- (
Reference:
Attachment 1) . 1 CORPORATE: Corporate Office is located in Juno Beach, Florida approximately 40 miles
- south of the plant on U.S.1.
(
Reference:
50 mile EPZ Environmental Monitoring Map) , State EOC: State of Florida Emergency Operations Center is located in Tallahassee, Florida. The State Warning Point (the 24 hour contact point for the Division 1 , of Emergency Management) is collocated with the EOC.
- County EOC: The St. Imcie County Emergency Operations Center is located approximately L 15 miles north west of the plant just off of Orange Ave. The St. Lucie County EOC is a hardened facility per FEMA guidelines.
(
Reference:
10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) i h ' 'Ibe Martin County Emergency Operations Center is located approximately 15 miles south of the plant just off of Port Salerno Rd. The Martin County EO is i a hardened facility per FEMA guidelines. (
Reference:
10 mile EPZ Environmental Monitoring Map for the SL Lucie Plant) ; r n . F ( s i
July 24,1992 j SITE & POPULATION Site 14 cation: St. Lucie site is located on Hutchinson Island, FL Approximately 71/2 miles south of Ft. Pierce, FL (10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) Coordinates: IAT 20:20'55" N (Unit 1) LONG 80:14'47" W (Unit 1)
, IAT 27:20'58". N (Unit 2) LONG 80:14'48" W (Unit 2)
(
Reference:
FS/R) Description of St. Lucie is a coastal plant (barrier island) with an eastern boundary of the the Environs: Atlantic Ocean and a western boundary of the Indian River. Population Distribution (Max population, resident and transient): 2-Mile Ring 658 5-Mile Ring 18,846 10 Mile Ring 124,019 Source: HMM document No 2336-1/ENV/706 Imcal Conununities within the.10 mile EPZ: Ft. Pierce Port St. Lucie Stuart West Palm Beach States wi. thin the 50 mile EPZ: Florida l Nearest ropulation Center (gmater than 25,000): i Ft. Pierce (5-10 miles) } Port St. Lucie (3-13 miles) ! i I
' Source: HMM document No 2336-1/ENV/706 i
- I I
2 Saint Lucie Units 1 & 2, REV 0
- July 24,1992 EMERGENCY RESPONSE OFFICL4LS 4 ,
2 . Licensee Representative with Authority to Make Protective Action Recommendations: Prior to staffing the Emergency Response Organization: ( l Nuclear Plant Supervisor on shift in the emergency position of Emergency Coordinator. -
) ! Upon arrival in the Control Room and after proper turnover, the Plant Manager (or } other senior management member) can take over as the Emergency Coordinator. ! After Staffing the Emergency Response Organization: ! The Plant Manager (or other senior management member) can take over as the i '
Emergency Coordinator in the Control Room or the TSC. ] i When the EOF is operational and a proper turnover is given, the Recovery Manager e assumes the responsibility for protective action recommendations from the Emergency Coordinator onsite. He also assumes offsite (state and local) notifications. ! Licensee Representative with Authority to Make In-Plant Technical Recommendations: ~ Nuclear Plant Supervisor on shift in the Control Room functioning as the Emergency Coordinator. ~ Upon arrival in the Control Room and after proper turnover, the Plant Manager (or
- other senior management member) can take over as the Emergency Coordinator.
- l
- After Staffing the Emergency Response Organization: l l De Plant Manager (or other senior management member) functioning as the !
Emergency Coordinator in the Control Room or the TSC. ! j Wheu the EOF is operational, the Recovery Manager is the senior manager in charge of {; the emergency onsite. State / Local Government Representative with Authority to Make Protective Action
, Decisions:
1 Initial Stages prior to Governor signing an executive order and State and locals in their own
- EOC (not in EOF):
He highest ranking member of county management in the EOC (preferably the l
- Chairmen of County Comminion) with input from the State and DHRS Office of Radiation Control (Orlando). -
\
4 . 3 Saint Lucie Units 1 & 2, REV 0 I
1 July 24,1992 After the Governor signs an executive order and State and locals in their own EOC (not in ' EOF): The Governors Authorized Representative in the State EOC with input from the State DHRS Office of Radiation Control and local government. t Prior to Governor signing an executive order and State and locals are in the EOF: . i The county emergency management director in the EOF with input from the State GAR and DHRS Office of Radiation Control. 1 After the Governor signs an executive order and the State and locals are in the EOF: i
;. The GAR in the EOF with input from DHRS Office of Radiation Control and local j government. ,
i l
- l 4
i [ i i 1 I i j
. l i
i l 4 Saint Lucie Units 1 & 2, REV 0
S'ECTION B APPENDlX: RESPONSE DRAWINGS, CHARTS AND MAPS TABLE of CONTENTS Site Plan , Emergency Planning Zone Sector Map Evacuation Routes Licensee Emergency Organization ! State Protective Action Decision Matrix i i d
- - NRC Plant Informatlos Book
- I SITE PLAN n
I u 9 l 1 i l 1 I w' J t 4 4 t 4 S e M e e 9 0
FIGURE 2-6 ST. LUCIE PLANT EMERGENCY FACILITIES LOCATION MAP l I 5 i
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EVACUATION ROUTES d 4 i i i I i 1 s 4 4 ' ? d Q* i I 4 u f l e ? .i 1 l i l 1 a b , J. ', "- l ! l 4 ! . ) l b
nhwnuur s FIEURE 5-2 GENERAL PUBLIC EVACUATION ROUTES N
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St. Lucie Rev.26 1
- - a 's* A a -e - a w . - -,- ,
H- p m = 6-- m a l t l , l t 4 4 4 e i ? a EMERGENCY PLAXNIN'G ORGANIZATION l .
, 9 i
f J b i < l l l e E
, i 4
4 e d 9 9 e e i 1 8-e s *
- r
m * - T 9 - 6 . VICE PRESIDENT-ST.LUCIE PUWT CD i i l i i i d
- SITE QUALITY HUMAN RESOURCES PUWT GENERAL PUWT UCENSING SERVICES b MANAGER MANAGER s* MANAGER MANAGER MANAGER 9 m
' T OUAuTvCONTROL _ TRAMWG z .
SUPERVISOR MANAGER -4 Z i I I I o 3 MAINTENANCE OPERATIONS TECHNICAL PROTECTION sp MANAGER MANAGER MANAGER h3 rc g < N 8 (, i l l om m _ HEALTH PHYSICS SHIFTTECHNICAL PREP mM SORS SUPERVISOR SOR COORDWATOR f 6' d . I Z NG SHIFTTECHNICAL , E _NGSOR , O CHEMISTRY Z SUPERVISOR y ' d o OPERATIONS Z SUPERVISOR I co
- Reports to Vkm Preeldent. Nudeer Assurance (EPUW4WPG) g SUPERVISOR 3; Ea I fv,
. . . _ , . _ . . . . ________.______.___.__.___m.____. _ _ _ . _ . _ .______m_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ . - . _ , _ _ . _ _-1
Arr$tuHsurAG M"URE 24
- IMMEDIATE RESPONSE ORGANIZATION NUCLEAR PLANT SUPERVISOR /
EMERGENCY. COORDINATOR 1 SHIFT ASSISTANT TECHNICAL NUCLEAR PLANT ADVISOR SUPERVISOR I I INTERIM UCENSED PLANT NUCLEAR WATCH RADIATION TEAM ENGINEER OPERATORS LEADER l 1 NON-UCENSED
- FIRE PUWT TEAM LEADER OPERATORS INTERIM l
FIRSTAID/ - DECONTAMINATION i TEAM LEADER INTERIM SECURIT/ TEAM LEADER ~(EPLN24.WP) J EP3:4 2-19 St. Lucie Rev.26 J
MEURE 2-1 FPL EMERGENCY RESPONSE ORGANIZATION a
- EMERGENCY EMERGENCY COEROL , INFORMATION OFFICER MANAGER RECOVERY EMERGENCY GOVERNMENTAL M MANAGER SECURITY -
AFFAIRS LL MANAGER MANAGER LL -
. o i-
! EMERGENCY TECHNICAL MANAGER
., )
NUCLEAR EMERGENCY PLANT GENERAL PLANT COORDINATOR . MANAGER OR 1 SUPV. (EC) ALTERNATE l i l INTERIM TEAM LEADERS EMERGENCY TEAMS PRIMARYTEAM LEADERS g b- u,,, Recmp&~ As Direchd by u) Fr -. Recomty Manager Z O S.nior Heshh Ppymes
- Radiation Noahn pny.m.
Supennsor FW .-_- - orAllemam . Nuclear Watch Engineer Nuclear Weten Engineer Ahomate Fire M Nudear Plant Oper. (NPO)
- Senior Nuclear Plant Operator or First Aid & Chem Supervisor Assocess Nuclear Plant Decontamination or Operator (ANPO) j' Securtty Shift Supervisor Plant Security Supervisor Altomate Security or Protected Area Guard Altemato (Formed as Required)
(EPt.N2-1.WPG) EP3:4 - 2-4 St. Lucie Rev.26
TABLE 2-2b FLORIDA POWER & LIENT EMER2ENCY RESPONSE OR2ANIZATION l l FUNCTIONS AND RESPONSIBILITIES - ! RESPONSIBILITY Function immediate Exoanded
. Command and Control . Emergency Coordinator Recovery Manager .
- (Nuclear Plant Supervisor) I
- . - Waming - Emergency; Coordinator- ,
Recovery Manager !' ; Notification Emergency Coordinator Recovery Manager
- j. Communications
]
! Public Information Emergency information Emergency Information i Manager - Manager
- J l Accident Assessment Emergency Coordinator 1_ Recovery Mana~ger (assisted !
n . (assisted by Shift Technical' by Emergency Technical l Advisor) Manager and his/her staff) l Fire Fire Team Leader . Fire Team Leader Rescue Interim Radiation-Team Primary Radiation Team l Leader Leader i Traffic Control interim Security Team Primary Security Team (on-site) Leader Leader . Emergency Medical Interim First Primary First Services ,__ Aid / Decontamination Team Aid / Decontamination Team Leader- Leader Transportation Interim Security Team Emergency Security Leader Manager
~
Protective Response ' Emergency Coordinator Radiation Team Leader
, (on-site) (assisted by RM's staff)
Radiological Exposure Emergency Coordinator Radiation Team Leader Control _(on-site) (assisted by RM's staff) Radiological Dose Emergency Coordinator Recovery Manager (assisted Assessment (assisted by Chemistry by Recovery Manager's Department representative) Staff) n EP3:4" - 2-29 St. Lucie
- Rev.26
.e
ATr4cwemy4EE Fl2URE 1-2 - INITIAL NOTIFICATION EMERGENCY i COORDINATOR Y STATE OF FLORIDA DMSION OF g EMERGENCY 4 A F DLRY CALL SUPERVISOR 7 PLANTGENERAL MANAGER MANAGEMENT w Y ; STATE OF FLORIDA DHR8 OrRC: OF RAoumON 4E -> EMERGENCY Tr4M LEADER r L r EMERGENCY TEAM uEMsERS 1 M NES NUCLEAR DMS40N
""E'E" iE 4- -> '
wrvorncea 4 rPtOrrStrE ogqECTORS
~
% ORGANIZATION EMERGENCY e /
USNRC SYSTEM OPERATION i OPERATIONS % > POWER COORDINATOR CENTER
. ~)!
)
SUPPORT INTEHIM
- 1. MREIAMSULANCF J EMNEW j P MEDCAL , r TEAMS ,
(1) Ma suas Hot R>g Down Tehofinne (Hfe) 1 PRM4ARY NOTIFICATON PATHWAY (2) Ms % N 9ymmm (ENS) ALTDNATE NOTIFICATION PATHWAY l (3) W & M Emmsganche enty,as needed ' s Me r=ePe=4a sramm i (5) N000 h to Emmgancy Omned OMker hr l Wed Nodheon l 4 4 EP3:4 1-14 St. Lucie Rev.26
F12URE 2-5 ! EXPANDED RESPONSE ORGANIZATION
\
4 l EMERGENCY CONTROL ! OFFICER l ) EMERGENCY RECOVERY GOVERNMENTAL INFORMATION A OFFICER MANAGER
~ , Q. MANAGER. ,
I I ]- EMERGENCY EMERGENCY - SECURITY TECHNICAL MANAGER MANAGER ] l l l l
~
EMERGENCY ADDNK STAFF AS COORDINATOR .
. REQUIRED (DFT6.WPG) a b ~
i ) t i EP3:4 2-34 St. Lucie Rev.26 a
. --..-. . . _ _ . . - .. ~ . . _ . _ - .
ATTACHrWD8T* GA , FIEURE 2-2A STATE, LOCAL, AND FEDERAL RESPONSE BEFORE EXECUTIVE ORDER Attomey General Govemor Lloonsee j , Department of Community Affairs Risk Counties
- Division of Emergency Management I Legend Direction Host Counties Coordination Otheringestion
, Fvpannte Pethway Counties I
Departmentof Health Department of Department Red d I Environmental and Rehabilitative Cross of LAW ' Services Regulation Enforcement Transpoen j p J Highway Safety & Motor Vehicles h ral h m p. ! es
^
.. . . l e p4 4
Department of # ' g g Military Affairs Consumer h Game & Fresh Water Fish Commission l (DFT2WPG) J EP3:4 2-5 St. Lucie ' Rev.26 4
- FISURE 2-23 STATE, LOCAL, AND FEDERAL RESPONSE AFTER EXECUTIVE ORDER i
Attomey General Govemor Uconsee Department of Community Affairs , Division of Emergency Management l i Risk Counties Hoot Counties Otheri Exposure athway Coordination
- j Depanmentof Health Department of Departnwnt l Red Department of and Rehabilitative Environmental Cmes W* transportation !
Services Regulation Enforcement Depanment W Department of NWNA Natural Resources Motor vehicles j ' ~ t-Department of d A0riculture & Military N Consumer Services Ofsh Com (OFD.WPG) 4 l 4 EP3:4. 2-6 St. Lucie Rev.26 4 6 g , - e s-
h i 4 a i STATE PROTECTIVE ACTION DECISION MATRIX h I i . h l l 1 4 t 4 1 1 4 i e f 8 4 4 M l J G v.,-
, PIGURE R-3 PRIMARY AND SECONDARY RESPONSIBILITIES l ST. LUCIE. COUNTY J l 1 s m E H a g lt lI W k, < g 8 E l I R ! 8 ; i h i g s, g I s. l l p
. e l
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" 5 8 "E 88 5 1
a E d N *- 5 ' E i m U E 4 E R E C H E E u 3 5 ; DTBBCTIOg un eaMTROL P P P i EMERGENCY ALERT AND MOTIPICATION P
- ea==UNIcwIa==- P S i ACCIDENT " ESaunuT S S P
- 1 PROTECTIVE masPnMBE P P S PUBLIC 17WmT mun NOTIPICATION P S S S S
!. PUBLIC IMPORMATION P S S 4 m anIOfMI"17. ETDOSURE CONTROL P S + DECOMTAMIMATION P S j CONTROL OF ACCESS TO EVACUATED ARM 1 S P S i PIELD MONITORING AMD E m LING S P S l PIRE un am8 CUE P l munnGEMCY unnIcar SERVICES P g LAW M ORC u muT P 4 TRANSPORTATION 8 j POOD OUALITY S P 4 POTABLE WATER QUALITY S P } SMELTER /CAnn 5 e , PUBLIC BR17TM AMD SAMITATION P SOCIAL SERVICES
; ROAD PASSAGE AND MAINT m _MCE SECURITY P TRAf7IC CONTROL P S RECOVERY AMD REENTRY ,
S S P l l: 4 . R-57 revision 12/31/92
i PIGURE R-5 ) l PavunnY AND SECONDARY RESPONSIBILITIES MARTIN COUNTY I. l 4 v E l E 8 N j . M" m< > o E 4 - m g _ g e o ! f I= t i I l "B 8 < a " 5 $ g ! 8 i: H o E E 8 9! 5 , ca - E z - - l l l E UR E d ! ! u l DIRECTION nun CONTROL P P S nuunGauCY 172nT un NOTIFICATION P
'anMMUNICATIONS P S AOCIDENT ***Egewmuy a S P P PEUTEdAva **mPOMSE P P 8 PUBLIC 172*T un MOTIFICATION P S S
, PUBLIC IMPORM1 TION P S S nanIOIDGInnt EXPOSURE CONTROL P DECOMT1mTu1 TION P S S S , CONTROL OF Acomum TO EVACUATED _ anum 3 p 3
- FIELD MONITORTMG aun 31msLTuG g P s FIRE OPERATIONS P unRGENCY MEDIN 1L OPERATIONS P LAW ENFORCrunT P TRANSPORTATION P
, FOOD OURLTTY P' S POThBLE WATER OUALITY 8 P SHELTan/enne g p g e
, PUBLIC WW1_LTM AMD S_1NITATION P , e CEs s ROAD PASSAGE AMD MAINTENANCE P SECURITY P TRAFFIC CONTROL P RECOVERY AMD REENTRY , 8 8 8 8 P R-59 revision 12/31/92
^
l i-f FIGURE R-7 1
- ymTummY AMD. SECONDARY RESPONSIBILITIES I INDIAN RIVER COUNTY .
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M $ $ f . 4 Diu-diOM mun enuTROL F P i mumma=nsey ar.mmT mMn MDTIFIe1'FION P S { navarriarIc1TInnen P l A n e r n ensT _m m e g g m u nnsy 7
- j. PROTECTIVE masPnismE F S j PUBLIC 1rMBT Rafn MOTIFICATION P S S S
! PUBLIC TuTO*n81 TION P P { manIOrmremr- Ex-usums CONTROL S P S S nanansThuT181 TION ' 1 P S S - nnesTEOL OF Aa"*** TO EY1cDATED 1881 ! PIntn unwrTOmfug mun amusLTuG B P ! PIRE mun m***UE P ? munnamnseY unnInst SERVICES S P
- tan marwomemmassT P l TRAMSPORTATION P S S 3 Pnnn OtttLTTY S P POTABLE-WATER OURLTTY S P
, a m e L e n a / e__m m e P S S S S
- i PumLTc am1LTM marn amarITATION P S SOcfmL Samviema 7
!- Sann Pasm1GE anon MRTMTansmuseE P l SECURITY P j' T***PIc CONTROL P RECOVERY AND REENTRY P S I . l R-61 revision 12/31/92 4
.,,,,e,,,, ,,,,,-,-,~w,. --...--,n- -
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. PIGURE R-9 PRIMARY.AND SECONDARY RESPONSIBILITIES PALM BEACE COUNTY J 4
i b i s s g G s s . E & l! I ai l si s s e - l p~ g e
- i i
- E i: E g 8 5 i __
l ii i l li li i DIRECTION AMD CONTROL P P i EMERGENCY ALERT Aun NOTIFICATION P COxuUmICATIONS P S ACCIDENT ASSESAMMMT P l PRuiEGiAva amePONSE P S S
- PUBLIC 17.2mT mun NOTIFICATION l PUBLIC Tuponw1 TION P S manIOLN3 Intr. ersOSURE CONTROL 5 P DECONTimru1 TION P
- CONTROL OF ACCESS TO EVACUATED ARE1 FImr.n unufTORING AMD SAMDLING FIRE AMD RESCUE i MMERGENCY MenInst. SERVICES S P LAW ENFORCRMENT P TRAMSPORTATION 8 P FOOD OUALITY P POTABLE WATER ODALITY P SHELTER / CARE P S S S
- PUBLIC REALTR AND SANITATION SOCIAL SERVICES S P ROAD PASSAGE AMD MAINTENANCE P ,
SECURITY P I TRAFFIC CONTROL P
- RECOVERY AND REENTRY ,
i 8 l R-63 revision 12/31/9: i
FIGURE'R-11 1 ' 1 PRIMARY AMD SECONDARY RESPONSIBILITIES I BRETARD COUNTY t 1 l
- l l
b 85 . ! IEl l h w l
!i a
- $9 =
c e 8 la - m g . m e
- a g " b g B E i
- !! E 5 d e i
IR IdH E E E 1 DIRECTION AMD CONTROL P P MMW8'2ENCY ALERT Dun NOTIFICATION P S ! y- RusiXCATIONS P ACCIDENT ASSESSMENT P 1 PROTECTIVE REEPONSE P S l 9 PUBLIC m m T m NOTIFICATION P l PUBLIC INFO m TION P 81MIOLOGInt m OSURE CONTROL P S l DECONTAMINATION P S CONTROL OF ACCESS TO EVACUATED AREA P _ FIntn MONITORING aun SamsLING P S P FIRE AND RESCUE P EMERGENCY MEDICAL SERVICES P LAW ENFORCEMENT P , TRANSPORTATION P l FOOD OUALITY P P ! POTABLE WATER OUALITY P P SHELTER / CARE S P S
- PUBLIC NRALTE AND BAMITATION P SOCIAL SERVICES P ROAD PASSAGE AMD MAINTENANCE P SECURITY P TRAFFIC CONTROL P RECOVERY AND REENTRY S P -
i I
, R-es f
revision 12/31/92
- i. .
{ IAYI (g 5'f) ' l B. EMERGENCY RESPONSE INFORMATION 1 i j SAINT LUCIE UNITS 1 and 2 :
- 1 TABLE of CONTENTS
)
4 j Emergency Response Facilities i Site and Population Emergency Response Officials Appendix: Drawings, Charts and Maps 4 1
. .a 4
d i NRC Plant leformados Book -
l l i i I I B. EMERGENCY RMPONSE INFORMATION i Plant Name: St. Lucie Unit 1 & 2 - 1 EMERGENCY RESPONSE FACHIRRS , . TSC: Technical Support Center is located adjacent to the Unit 1 Control Room at the St. Lucie Plant. Reference Attachment 1. EOF: Emergency Operations Facility is located at thd Midway Substation ; i approximately 10.1 miles due west of the power plant at the intersection of l state Road 712 and I-95. ' (Riw 10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) l l 4 i . LOSC: Operational Support Center is located in the North Service Building in the i Conference Room on the second floor. i { (
Reference:
Arrachment 1)
~
CORPORATE: Corporate Office is located in Juno Beach, Florida approximately 40 miles south of the plant on U.S.1. .
! '(
Reference:
50 mile EPZ Envira== ental Monitoring Map) l State EOC: ' State of Florida Emergency Operations Center is located in T=11=h===aa Florida. The State Warning Point (the 24 hour contact point for the Division { j , of Emergency Management) is collocated with the EOC. t County EOC: The St. Lucie County Emergency Operations Center is located approximately t
' 15 ' miles north west of the plant just off of Orange Ave. The St. Lucie County EOC is a hardened facility per FEMA guidelines. , (
Reference:
10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) " The Martin County Emergency Operations Center is located approximately 15 miles south of the plant just off of Port Salerno Rd. The Martin County EO is
, a hardened facility per FEMA guidelines.' .
(
Reference:
10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) 9 4
~
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. i + e 3
July 24,1992 SITE & POPULATION Site Imcation: St. Lucie site is located'on Hutchinson Island, FL Approximately 71/2 miles south of Ft. Pierce, FL (10 mile EPZ Environmental Monitoring Map for the St. Lucie Plant) Coordinates: LAT 20:20'55" N (Unit 1) LONG 80:14'47" W (Unit 1) LAT 27:20'58" N (Unit 2) LONG 80:14'48" W (Unit 2) (
Reference:
FSAR) Description of St. Lucie is a coastal plant (barrier island) with an eastern boundary of the
' the Envimas: Atlantic Ocean and a western boundary of the Indian River.
Population Distribution (Max population, resident and transient): 2-Mile Ring 658 5-Mile Ring 18,846 10 Mile Ring ' 124,019 Source: HMM document No 23361/ENV/706 Iscal Comununities within the 10 mile EPZ: Ft. Pierce Port St. Lucie Stuart West Palm Beach 4 States within the 50 mile EPZ: j Florida ' Nearest Population Center (greater than 25,000): Ft. Pierce (5-10 miles) Port St. Lucie (3-13 miles) Source: HMM document No 2336-1/ENV/706 2 Saint Lucie Units 1 & 2, REV 0
, ' July 24,1992 EMERGENCY RESPONSE OFFICIALS 1
Licensee Representative with Authority to.Make frotective Action Recommendations: Prior to staffing the Emergency Response Organization: Nuclear Plant Supervisor on shift in the emergency position of Emergency Coordinator.
- Upon arrival in the Control Room and after proper turnover, the Plant Manager (or other senior management member) can take over as the Emergency Coordinator.
j After Staffing the Emergency Response Organization: 4 ! The Plant Manager (or other senior management member) can take over as the Emergency Coordinator in the Control Room or the TSC. ] ! When the EOF is operational and a proper turnover is given, the Recovery Manager i
' assumes the responsibility for protective action recommendations from the Emergency Coordinator onsite. He also assumes offsite (state and local) notifications.
3 IJcensee Repmsentative with Authority to Make In-Plant Technical Recomunendations: l Nuclear Plant Supervisor on shift in the Control Room functioning as the Emergency j ~ Coordinator. i i Upon arrival in the Control Room and after proper turnover, the Plant Manager (or ] other senior management member) can take over as the Emergency Coordinator. l After Staffing the Emergency Response Organization: l The Plant Manager (or other senior management member) functioning as the Emergency Coordinator in the Control Room or the TSC. When the EOF is operational, the Recovery Manager is the senior manager in charge of the emergency onsite. l , l State / local Government Representative with Authority to Make Protective Action j Decisions: i
- Initial Stages prior to Governor signing an executive order and State and locals in their own EOC (not in EOF)
- 1 1
'Ibe highest ranking member of county management in the EOC (preferably the 1 Chairmen of County Comminion) with input from the State and DHRS Office of Radiation Control (Orlando). .
, 1 l 1 , l 3 Saint Lucie Units 1 & 2, REV 0 __ . .. __ .. - - -_ - .-_1
July 24,1992 l After the Governor signs an executive order and State and locals in their own EOC (not in EOF): 4 The Governors Authorized Representative in the State EOC with input from the State l . DHRS Office of Radiation Control and local government. Prior to Governor signing an executive order and State and locals are in the EOF: The county emergency management director in the EOF with input from the State GAR and DHRS Office of Radiation Control. { After the Governor signs an executive order and the State and locals are in the' EOF:
; The GAR in the EOF with input from DHRS Office of Radiation Control and local government. ,
I 3 i 1 4 5 4 Saint Lucie Units 1 & 2, REV 0
a SECTION B APPENDIX: RESPONSE DRAWINGS, CHARTS'AND MAPS i 4 TABLE of CONTENTS 4 Site Plan Emergency Planning Zone Sector Map Evacuation Routes Licensee Emergency Organization State Protective Action Decision Matrix 1 l l 1 NRC Plant Information Book
I 4 SITE PLAN 4 9 d i 1 A s
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, ST. LUCIE PLANT 4
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Warning Emergency Coordinator Recovery Manager Notification Emergency Coordinator' Communications Recovery Manager I l Public Information Emergency Information Emergency Info. i Manager Manager l Accident Emergency Coordinator Recovery Manager 4 Assessment (assisted by Shift (assisted by i Technical Advisor) Emergency Tech-l- nical Manager and his staff) Fire Fire Team Leadar Fire Team Leader
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- Emergency Team Leader Security Manager k Protective Response Emergency Coordinator (on-site) Radiation Team Leader (assisted by RM's staff) Radiological Emergency Coordinator Exposure Control Radiation Team (on-site) Leader (assisted
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Radiological Dose , Emergency Coordinator Recovery Manager Assessment (assisted by Chemistry p Department representative) l [ EP3:4' 2-25 St. Lucie Rev. 21 !
i NM Page a of 24 g ST. LUCIE PLANT b ' 3 y E-PLAN IMPLEMENTING PROCEDURE NO. 3100023E, REVISION i ON-SffE EMERGENCY ORGANIZATION AND CAII DIRE - t i
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4 OA : dwb4 i C. PLANT' DESCRIPTION
SUMMARY
I SAINT LUClE UNIT 2 l l l i 4 i 4 i NOTE: This Section was reproduced from Chapter 1 of the Final Safety 2 Analysis Report (FSAR) , 1 5 1 i I i 1 NRC Plant Infonnation Book .
=
1.2 CENERAL PLANT DESCRIPTION i 4 1.2.1 PRINCIPAL SITE CHARACTERISTICS l 1 The site for St Lucia Units 1 and 2 consists of approximately 1,132 acras. The unimproved area of the site is generally flat, covered witn water and has a dense vegetation characteristic of Florida coastal man- l l grove swamps. At the ocean shore the land rises slightly in a dune or l ridge to approximately 15 ft. above mean low water. l I The island and the adjoining mainland are sparsely populated. The southern I most boundary of the nearest population center is the City of Fort Pierce which is 4.1 miles from the site. The City of Fort Pierce has an estimated population of 33,083 people as of a 1978 estimate. The minimum site i exclusion radius is 5,100 ft. Site characteristics are given in Chapter 2. I 1.2.2 PRINCIPAL DESIGN CRITERIA . Principal structures, systems and equipment which may serve either to i prevent accidents or to mitigate their consequences arn designed and ' l erected in accordance with applicable codes to withstand the most severe
- earthquakes, flooding conditions, windstorms, temperature and other deleterious natural phenomena which could be reasonably assumed to occur at the site during the lifetime of the plant. Principal structures, systems and equipment are sized for the design power level of the nuclear steam supply system output.
Redundancy is provided in the reactor protective and engineered safety feature systems so that no single failure of any active component of the systems can prevent action necessary to avoid an unsafe condition. The plant is designed to facilitate inspection and testing of systems and components whose reliability are important to plant shutdown and to the , , protection of the public and plant personnel. Provisions are made to minimize the probability and effect of fires and
- explosions, in accordance with 10 CFR 50, Appendix R.
1 Systems and components which are significant from the standpoint of nuclear safety are designed, fabricated and erected to quality standards cosunen-
- surate with the safety function to be performed.
Section 3.1 addresses the implementation of the NRC General Design Criteria for Nuclear Power Plants, 10 CFR Part 50, Appendix A. Chapter 17 describes the quality assurance program for the design and operation of St. Lucia Unit 2. 1.2.2.1 Reactor The reactor is of the pressurized water-type, designed to provide heat to senam generators which, in turn, provide stama to drive a turbine Renerator. The full power core thermal output is 2700 megawatts. l1 The reactor core is fueled with uranium dioxide pelle.cs enclosed in zircaloy tubes pressurized with helium and fitted with welded end plugs. 1.2-1 Amendment No. 1, (4/86)
6
' The tubes are fabricated into assemblies in which end fittings prevent I axial motion and spacer grids ' prevent lateral motion of the tubes.- The i control element assemblies (CEAs) consist of inconel clad boron carbide absorber rods which are guided by zircaloy tubes located within the fuel assembly. The core consists of 217 fuel assemblies with three U-235 enrichments in a three batch, mixed central zone arrangement.
i Minimum departure from nucleate boiling ratio (DNER) during normal opers-I tion and anticipated operational occurrences is not less than 1.28 (cycla 3 I was 1.19) using the CE-1 correlation. The maximum center line temperature i of the fuel, evaluated at the design overpower condition, is below that ] value which could lead to fuel rod failure. The melting point of the UO2 is i not reached during routine operation and anticipated operational occurrences. 1 4 The combined response of the fuel temperature coefficient, the moderator j temperature coefficient, the moderator void coefficient and the moderator 1 i pressure coefficient to an increase in reactor thermal power is a decrease l j in reactivity. In addition, the reactor power transient remains bounded l and damped as response to any expected changes in any operating variable. 4 control element assemblies (CEAs) are capable of holding the core sub-critical at hot sero power conditions with margin following 's trip. even j with the most reactive CIA stuck in the fully withdrawn position. j Fuel rod clad is designed to maintain cladding integrity throughout fuel life.- Fission gas release within the rods and other factors affecting
- design life are considered for the maximum espected exposures.
The reactor and control systems are designed so that any xenon transients { are adequately damped. l The reactor in conjunction with the Reactor Protective System is designed j to accommodate, safely and without fuel demage. the anticipated operational j occurrences. i The reactor vessel and its closure head are fabricated from manganesemoly steel. internally clad with austenitic stainless steel. The vessel and its } internalgare dgsigned. so that the integrated neutron flux does not exceed t 3.2 x 10 n/cm (E > l Nev) over the 40 year design life of the vessel. l i ' ' Power excursions ubich could result from any credible reactivity addition do not cause damage, either by deforestion or rupture of the reactor vessel , i
- and do not impair operation of the Engineered Safety Features.
The internal structures include the core support barrel, the lower support i structure, the core shroud, the hold-down ring and the upper guide structure assembly. The core support barrel is a right circular cylinder supported. I from a ring flange from a ledge on the reactor vessel. The flange carries the entire weight of the core. The lower support structure transmits the weight of the core to the core support barrel by means of vertical columns and a beam . structure. The core shroud surrounds the core and limits the j amount of coolant bypass flow. The upper g' u ide structure provides a flow j shroud for the CEAs and prevents upward motion of the fuel assemblies j during pressure transients. Lateral motion limiters or snubbers are i i 1.2-2 Amendment No. 1, (4/86)
provided at the-lower end of the core support barrel assembly. The hold-down ring acts as a shim. and is set between the reactor vessel head and the upper guide structure to resist axial upward movement.- Further details concerning. the reactor are given in Chapters 3 and 4
-1.2.2.2 Reactor Coolant and Auxiliary Systems The Reactor Coolant System is arranged as two closed loops connected in parallel to the reactor vessel. Each loop consists of one 42 in. ID outlet (hot) pipe, one steam generator, two 30 in. ID inlet (cold) pipes .
and_two reactor coolant pumps. An electric' ally heated pressuriser. is connected to the hot leg of one of the loops and a safety injection line I is connected to each of the four cold legs. { l The Reactor Coolant System operates at a nominal pressure of 2235 psig. The reactor coolant enters near the top of the reactor vessel, and. flows e; downward between the reactor vessel shell. and the core support barrel. ' into the lower plenum. It then flows upward through the core. leaves the i reactor vessel, and flows through the tube side of the two vertical U-tube l steam generators where heat is ' transferred to the secondary system. ' Reactor coolant pumps return the reactor coolant to the reactor vessel. The two steam generators are vertical shell and U-tube units. The steam generated in the shell side of the steam generator flows upward through moisture ' separators and scrubber plate dryers which reduce the moisture l content to less than 0.2 percent. All surfaces in contact with the reactor coolant are either stainless steel or NiCrFe alloy in order to minimize corrosion. . { The-reactor coolant is circulated by four electric motor driven single-suction vertical centrifugal pumps. The pump ' shafts are sealed by mechanical seals. Each pump motor is equipped with an antireverse mechanism to prevent reverse rotation. 4 Components of the Reactor Coolant System.are designed and operated so that no stresses are imposed on the structural materials that result in loss of . function. The necessary con' sideration has been given to the ductile charac-i teristics of the materials at low temperatures, The Resctor Coolant. System is designed and constructed to maintain its ; ! integrity throughout the plant life. Appropriate means of test and l 4 inspection are provided. , see Chapter 5 for further information.
'1.2.2.3 Engineered Safety Features i
, The plant design incorporates redundant Engineered Safety Features. These l [ systems.in conjunction with the containment system, ensure that the o f f- l site radiclogical consequences fo'11owing any LOCA up to and including a double ended break of the largest reactor coolant pipe, do not exceed the guidelines of 10CFR100. The systems also etnsure that the guidelines j of 10CTR50. Appendix K, " Acceptance Criteria for Emergency Core Cooling 1.2-3 . l
' Systems" are ' satisfied, based upon analytical methods, assumptions and proceduies. accepted by the NRC. ' The Engineered Safety Features include:
(a) independent redundant systems (Containment Cooling Sy, stem and Containment Spray System) to remove heat from and reduce the pressure in the containment vessel in order to maintain containment integrity, (b) a high and lov ' pressure Safety Injection System to limit fuel and cladding damage to an amount which does not interfere with adequate emergency core cooling and to limit metal-water reactions to negligible amounts, (c) a Shield Building Ventilation System and an Iodine Removal System to reduce offsite consequences 'due, to leakage from the containment vessel, (d) a containment isolation system to minimize post-LOCA radiological effects-of fsite , (e) a hydrogen control system to maintain safe post-LOCA hydrogen concentration within the containment, and (f) a control room habitability systes. The Reactor'Euilding, which is a dual containre'nt design', is comprised of a steel containment vessel surrounded by an annular space and enclosed by a reinforced concrete Shield'Euilding. The containment vessel is a-low leakage stwel'shell which is designed to confine the radioactive raterial that could be released from a postulated design basis, Loss-of-Coolant Acci-dent, (LOCA), resulting in release of fission products as' defined in. TID 14844 It is a cylindrical vessel with heeispherical dose and ellip-soldal bottom. The Shield Building is a. medium leakage concrete structure which surrounds the annulus and steel containment vessel. It containment vessel free external missiles, .and provides biolog'ical protects the shielding and a means of collecting radioactive fission products that may leak from the containment following a major hypothetical accident (see Subsection 6.2.1 for details). The containment in conjunction with either of the associated spray and cooling systems is designed to withstand the internal' pressure and coin-cident temperature resulting from the energy release associated with the design basis accident. The containment is equipped with two 100 percent capacity heat removal systems, each comprised of one containrent spray loop and two containment cooling units. The Containment Spray System supplies borated water to cool and reduce , i pressure in the containment staosphere. The pumps take suction initially i ! from the refueling water tank. Long term cooling is based on suction from the containment sump through the recirculation lines. l l The Containment Cooling' System provides containment atmosphere eixing by recirculation. The cooling coils and fans of the Containeent Cooling j
- System are sized to provide adequate containment cooling at post-accident ;
j conditions of terperature, pressure and humidity (see Subsection 6.7.2 for l details). ~ l i
- In the event of.a LOCA, the Safety Injection Systes described in Section l
! 6.3 injects borated water into the Reactor Coolant Systesi. This provides
- cooling to limit core damage And fission product release, and assures i i adequate shutdown margin. The injection systeni also provides continuous !
i 'long ' term post-accident cooling of the core by recirculation of borated l 1 , water f rom the containsent susp through the shutdown heat exchangers and 1
. ~ back to the reactor core. '
1.2-4
The Shield Building Ventilation System is provided to maintain a negative pressure . in the annulus between the steel containment vessel ' and the concrete Shield Building following a LOCA. 'Two independent 100 percent 4 capacity systems'are provided. This system filters any radioactivity leakage 'from the containment vessel and therefore reduces the effects ; on the environment (see Subsection 6.2.3 for details). The SBVS is provided with carbon adsorbers for iodine removal in the Shield' Building. ' The Iodine. Removal System is provided to enhance the capture of radio-iodines from the containment atmosphere . following a LOCA by adding con-crolled amounts of hydrasine to the containment spray water. Two in-dependent 100 percent capacity systems are provided (see Subsection 6.5.2 for details). A containment isola *. ion systen consisting of valves and associated actustors and controls is provided for each line penetrating the containment that must be closed to prevent a radioactivity release in the case of a loss-nf-coolant accident (see Subsection 6.2.4 for details). A hydrogen control system is provided which consists of redundant hydrogen ; recombiners and hydrogen sampling' systems. A hydrogen purge system is provided as a non-safety, diverse system in addition to the redundant
.recombiner system (see Subsection 6.2.5 for deta,ils). .
The control room habitability system is provided to limit control room doses from airborne activity to within CDC 19 11mics' (see Section 6.4 for details). . 1.2.2.4 Protection, Control. Instrumentation and Electrical Systems a) Reactor Protective System l The reactor parameters are maintained within the acceptable limits f
- by the inherent' characteristics of the reactor, by the Reactor
- Regulating System, by boron in the moderator and by the operating
! procedures. In addition. in order to preclude unsafe conditions for F plant equipment or personnel, the Reactor Protective System I initiates reactor trip if a selected parameter reaches 'its preset limit. Four : independent channels normally monitor each of the selected plant parameters. The Reactor Protective System logic initiates protective action whenever the signal of any two of three channels reaches the preset limit. A fourth channel is provided as a spare and allows bypassing of one channel while j maintaining a two-out-of-three system. If any two' channels receive L , coincident signals, the power supply to the magnetic jack control element drive mechanisms ic interrupted releasing the control elements , to drop into the core to shutdown the reactor. Redundancy is provided in the Reactor Protective System to assure that no single failure j1 1 prevents protective action when it is required. The protective } -system is completely independent of and separate from the control system (see Section 7.2 for details).
- l 1.2-5
. .- _ .~-
1 1 i l i l l b) Control System- j The reactor is controlled by a combination.of control element l assemblies (CEAa) and dissolved boric acid in the reactor coolant. 1 Boric acid is used for reactivity changes associated with'large l but aradual changes in water temperature, ' core xenon, fuel burnup
.and power levels. Additions of horic acid also provide an increased shutdown margin during the . initial loading and subsecuent refuel- ;
ings. The boric acid solution is prepared and stored at a temper- l sture'sufficiently high'to prevent precipitation. CEA movement j provides changes in reactivity for shutdown or power changes. The 1 CEAs are actuated by control drive mechanisms mounted on the ' reactor vessel head. The control drive mechanisms are designed to permit rapid insertion of the CFAs into the reactor core by . J gravity. CEA motion.can he initiated manually or automatically. I The Reactor Regulatine System (Subsection 7.7.1.1) provides for start up and shutdown 'of the reactor and for adjustment of the + reactor power in response to turbine load demand. The Puclear Steam Supply System is cap'able of following a ramp change from 15 percent to 100 percent pcwer at a race of five. percent 'per 1 minute and at greater rates over ses11er load change increments up to a step change of 10 percent. This control is ac,complished by movement of CEAs in response to a " change in reactor coolant j temperature. A temperature controller compares the existing average reactor coolant temperature with the value corresponding to the power called for by the temperature control program. If the temperature is different, the CEAs'are adjusted to bring the two temperatures within the prescribed control band. Regulation of the i reactor coolant temperature in accordance with this program maintains f' the secondary steam pressure within operating limits and matches reactor power to load demand. The pressure in the Peactor Coolant System is coctrolled by reFulat-l ing the temperature of the coolant in the pressuriser, where steam i and water are held in' tbermal equilibrium. . Steam is formed by the , j pressurizer heaters or condensed by the pressurizer spray to reduce ! variations caused by-expansion and contraction of the reactor , coolant temperature changes. The pressure and water level control I systems are described in Subsection 7.7.1.1. ) l Overpressure. protection of the Peactor Coolant System is provided by power operated relief valves and spring loaded safety valves con-I nected to the pres'surizer. The discharge from the pressurizer safety l and relief valves is released under water in the pressurizer quench tank, where it is condensed and cooled'. In the event'the' discharged steam exceeds the capacity of the tank, the tank relieves to the
- contatnment atmosphere (see Subsections' 5.2.2, 5.2.6, and 5.4.13 for l details). )
l
' l A Turbink Control System is provided to regulate steam flow.to the l l turbine as a function of system load. In,the event of turbine trip, '
bypass systems are provided to release steam to the condenser and to the' atmosphere. These systems are designed to reduce the sensible 1.2-u . 1 I w a . .
, . _D
4 heat in the Peactor Coolant System, maintain the steam renerator pressure during hot standby, and permit up to a 45 percent load rejection without opening the pressurizer safety valves, without opening the steam generator safety valves, 'or without causing reactor
; trip when the condenser is available (see Section 7.7)
A Steam Generator Water Level Control System regulates feedwater' flow to the steam generator (see Subsection 7.7.1.1). An Auxiliary Feedwater System is provided to ensure flow to the steam generators in the event the main feedwater supply is out of service. 4 c) Instrumentation System . The nuclear instrumentation includes excore and in-core neutron flux detectors. Twelve channels of excore instrumentation monttor the neutron flux and provide reactor protection and control signals during start up and power operation. Four of the channels are i f wide range logarithmic safety channels to measure neutron flux I d from* source range to above 200 percent of full power. Another four channels are power range safety channels to measure neutron flux linearly from one percent to 200 percent of full power. The power range safety channels are used by the reactor protection ! , system to determine the neutron flux power and axial offset, and by the high power bypass circuitry for the high rate-of-change l l of power trip (see Subsection 7.2.1.1). There are two linear power range channels utilized for control purposes and two channels for startup and extended shutdown (see Subsection 7.7.1.1.9). The in-core instrumentation consists of self powered rhodium neutron detectors and backaround detectors' to provide in' formation on neutron l f flux distributi'on. l l The process instrumentation monitoring. includes those critical chan-
- nels which are used for protective action. Temperature, pressure, flow and itquid level monitoring is provided, as reautred, to keep the operating personnel informed of plant conditions and to provide information from which plant processes can be evaluated and/or
, regulated. The boron concentration in the reactor coolant water j is also monitored and the concentration is displayed in the control room. 1 4 Instrument signals transmitted from the containevnt are electric. Instrument signal transatssion for the remai'ning plant. instruments 2 is either electric or pneumatic (see Chapter 7 for details). i The plant gaseous and liquid effluents are' monitored to assure that they are maintained within acceptable radioactivity limits. Activity levels are displayed and off-normal values are annunciated. Area' monitoring stations measure radioactivity at selected locations in the plant for personnel. protection. A complete description of F the radiation instrumentation is contained in Section 11.5 and Subsection 12.3.4 1.2-7
.- .- . .. -- - _ - . ~ - .- .. -. - _ . . - - .
. ~ .
l id) Electrical System j Redundant sources of offsite power are provided by three ' separate transmission lines. l The unit includes. a 1.000 MVA. 0.85 power factor generator
, delivering power to a 240 kV switchyard through step-up' power ]
transformers. Auxiliary power is utilized at 6.74 kV. (a 6.9 kV l winding is provided for the start up transformers) 4.16 kV. 480V. l and 120Y ac; 125V de systems are also provided. For emergency ~ l power. Engineered Safety Features'- control, and essential nuclear l instrumentation. all voltages except 6.74 kV are provided. . I The auxiliary load is normally supplied from two auxiliary trans-formers connected to the main generator bus. Start up power is supplied from two start up transformers connected to the 240 kV l switchyard. Emergency power for the Engineered Safety Features 1 is supplied by redundant diesel generator sets (see Chapter 8 1 for details). 1.2.2.5 Power Conversion System l , The. power conversion system removes heat energy, from the reactor coolant l l in two U-tube steam generators, and converts the steam into electrical ' energy by means of a turbine generator. The unusable heat in the steam , cycle is transferred to the main condenser for rejection by the Cir-
- culating Water System. The resulting condensate is deserated in the condenser. then heated through feedwater heaters and returned to the steam generators as feedwater.
l The turbine generator is a Westinghouse Electric Corporation unit. It is l i an 1.800 rpm tandem-compound, four-flow exhaust unit with 44 in last stage blades. The feedwater pumps are electric motor driven. Each of } two strings of feedwa'ter heaters consists of four low pressure a'nd one i high pressure heaters. 1 i The' Auxiliary Feedwater System contains two electric motor driven pumps and I one pump driven by a noncondensing steam turbine. This system provides a source of water inventory to the steam generators during plant startup and , hot standby and during plant cooldown provides heat removal to bring the j Reactor Coolant System to the shutdown cooling system activation window.
- . (See Chapter 10 for details.)
e 4 4 e e i 1.2-8
4 4 i
'1.2.2.6 . Fuel Handling and Storate Svetems ; ; The fuel handling systems provide for the safe handling of fuel assemblies ;
1 and control element assemblies under all foreseeable conditions and for {
- the required assembly, disassembly, and storage of the reactor vessel '
. head and internals. . These systems include a refueling machine located'inside containment above the refueling cavity, the fuel transfer carriage, the
- upending machines, the fuel transfer tube, a spent fuel handling machine in j the Fuel Fandling Building, and various devices used for handling the reactor vessel head and internals (see subsection' 9.1.4 for details).
i Pew fuel is stored dry in' vertical racks in the ruel Handling Pullding. t The rack and fuel nosembly spacine precludes crit'icality (see Subsection 0.1.1 for details). j The spent fuel pool is~ a reinforced concrete structure, stainlesi steel 1 Lned. . Spent fuel assemblies are stored in vertical racks so spaced as to l' preclude criticality with no credit taken for the boron in the pool water , j (see Subsection 9.1.2 for details).
- 4
[ Cooling and purification equipment is provided for the fuel pool water. 1 This equipment may also be used for cleanup of refueling water after each i fuel change in the reactor (see Subsection 9.1.3 for' details). l 1.2.2.7 Cooling Water and Ot'er h Auxiliary Systems
- - 1 4 .
l i a) Chemical and Volume Control System ' The purity level in the Reactor Coolant System is controlled by con-tinuous purification of a bypass stream of reactor coolant. Water j removed from the Reactor Coolant System is cooled in the regenerative t heat exchanger. From there, the coolant flows to the letdown heat exchanger and then through a filter and a domineralizar where corro- . sion and fission products are removed. It is the'n sprayed into the ! volume control tank and returned to the ,rerecerative heat exchanger 4 by the charging pumps where it is heated prior to return to the l Reactor Coolant System. The Chemical'and Volume Control System automatically adjusts the l amount of reactor coolant in order to maintain a constant level in the pressuriser. This-compensates for changes in specific volume i due to coolant temperature changes and reactor coolant pump shaft controlled seal leakage (see Subsection 0.3.4 for detalis). The Chemical and Volume Control System is capable of adding horic acid to the reactor coolant at a rate sufficient to maintain an , adequate shutdown marFin during Reactor Coolant System cooldown at the maximum design rate following a reactor trip.
~
The system is l
- independent of the CEA system.
- j l
- 6
, 4 1.2-9 .
.,_ __ . _ _ ~ _._ __ _. . _ _ _ _ - - _ _ _ . - _ . _ -
1 4 b) . IShutdown Cooling System 3 l The Shutdown Cooling System is used to reduce the temperature of the reactor coolant-at a controlled rate and to maintain the proper
' reactor coolant temperature during refueling.
4 The Shutdown Cooling System utilizes the low pressure safety injection s
. pumps to circulate the reactor coolant through two shutdown heat ex-j changers, returning it to the Reactor Coolant System through the low ] ! pressure injection header' (see Subsection 5.4.7) . 1 l The Component Cooling System serves as a heat sink for the shutdown l j heat exchangers.
c) Sampling Systes
, Two sampling. systems are provided; one for the reactor coolant and d
its auxiliary systems and one for the turbine steam and feedwater
- system. ' These systems are used for' determining both chemical and l
] radiochemical conditions of the various process fluids used in the l ) plant (see Subsection 9.3.2). 5 1 l d) Cooling Water Systems - l The turbine generator condenser is cooled by the Circulating Water j , System which takes suction from and discharges to the Atlantic Ocean.
. An Intaka Cooling Water Systes provides seawater from the Circulating Water System intake structure and serves as a heat sink for the com-j ponent cooling water heat exchangers, the Turbine Clos.ed Cooling
- System heat excharigers and the blowdown system open cooling water heat
- exchangers.
l The Component Cooling Water System, consisting of three pumps and two heat exchangers, removes heat from the various auxiliary systems.
~
- Corrosion inhibited desineralized water is circulated by the system 3 through auxiliary. components of the Nuclear Stema Supply System that
- require cooling water. During reactor shutdown, component cooling j water is .also circulated through the shutdown heat exchangers. The Component Cooling Water Systes provides an intermediate barrier be-l tween' the Reactor Coolant System and the Intake Cooling Water System j (see Subsection 9.2.2 for details).
The blowdown systes closed cooling water heat exchangers remove heat 1 from the steam generator blowdown. This heat is, in turn, removed by l the intake cooling water by the open blowdown cooling water system heat exchangers.
- The Turbine Closed Cooling Water System removes heat from the turbine
- . r,enerator oil cooler, hydrogen coolers, feed' pump oil coolers, sampte -
- j. , coolers, and other components' by providing corrosion inhibited de-
. mineralized water to those components {see Section 9.2 fer de-
, tails). .
1 i~ l 1.2-10 i , I
e) Plant Ventilation Systems Separate ventilation systems are provided for the containment vessel, the control room, the Reactor Auxiliary Building, the Fuel Handling i Building, Turbine Building, CCW structure, intake structure, and the { Diesel Generator Building. Two purge systems are provided for the containment atmosphere (see Section 9.4). The annular space between the steel containment vessel and the con-crete Shield Building is evacuated by the Shield Building Ven-tilation System utilizing charcoal filters.for. removal of radioactive ; iodine. This system is automatically put into . operation upon receipt of a containment isolation actuation signal following a LOCA (see Subsection 6.2.3).
' f) Plant Fire Protection-Systen The Fire Protection System, common to St Lucie Units 1 and 2, supplies water to fire hydrants, deluge systems and hose racks in the various areas of the plant. Additional design features are provided through-out the plant to ensure conformance to 10CFR50 Appendix A and Appendix R. (See Subsection 9.5.1 and Appendix 9.5A.)
g) Compressed Air System l l The Compressed Air System supplies properly conditioned compressed I air required to operate pneumatic instruments and controls, operate l containment isolation valv se and perform normal plant maintenance. It consists of the Instrument Air System which supplies the various air operated valves, pneumatic instruments and controls, and the Station Air System which supplies various outlets throughout the plant. Multiple compressor units and a cross-connection are provided between the Instrument and Station Air Systems. In case of loss of instrument air, all safety related pneumatically operated devices in the plant are designed to fail in a position which would allow safe, shutdown. 1 Where safety class valves are required to operate, accumulators are ] provided (see Subsection 9.3.1). i h) Diesel Generator Fuel Oil Storage and Transfer System 4 The Diesel Generator Fuel Oil System is provided to transfer diesel fuel oil from the onsite storage tanks to the day tanks which supply the emergency diesel generator sets. Redundant subsystems are pro-vided, capable of supplying sufficient fuel to their respective diesel generator sets. { e i + 1.2-11 i e e
,c - - -
l.2.2.8 Padioactive Waste Manaeement System The Waste Management System provides the means for controlled handling, storage and disposal- of liquid, gaseous and solid wastes. In addition, the system reconcentrates and recovers dissolved boron from the licuid effluent
.for reuse in the plant. The principal design criterion is that plant per-sonnel and the Feneral public are protected by ensuring that all normal operating releases of radioactive material are made as low as reasonably achievable in accordance with the provisions of 10CFR50, Appendix I.
Reactor. coolant from the Chemical.and Volume Control System and from the reactor drain tank is processed.by the boron recovery system, which is comprised of filters, flash tank, ion exchangers, concentrators, holding and condensate tanks. The concentrators are used to reconcentrate the boric
. acid. The concentrate is normally returned to the boric acid makeup tank in the Chemical and Volume Control System, but if the solution is unsuitable for reuse, provision is made to solidify and transport. it' from the plant to a disposal site. The distillate from the concentrator is sampled'and may be discharged to the Circulating Water System if the-radioactivity is within
- specified limits.
Miscellaneous liquid wastes from the Reactor Auxiliary Fullding are collected in the equipment and chemical drain tanks and subsecuently, processed by fil-f tration, ion exchanee and/or concentration. The distillate entere the vaste condensate tank. If the radioactivity level of the liquid in the condensate tank is found to be high, the vaste can be recycled through the waste ion i exchanger and waste concentrator. The llould in the holding tank is sampled l to ensure radioactivity levels are within the acceptable limits orier to dis-charge to the Circulating Water Fystem. The concentrate is containerized for i offsite disposal. Weste gases are' compressed and stored in the gas decay tanks which have a
- 30 day storare capacity. After decay, the gas in the waste gas decay tanks
- is sampled to ensure radioactivity levels are within acceptable limits, and j- is then released to the. plant vent at a controlled rate.
Spent Ion exchange resins and filters are ultimately transported in a i shielded container from the plant. i i Low activity wastes such as contaminated laundry, rags and paper are compact- ) ed and containerized for removal from the plant (see Chapter 11 for details). 1.2.3 MAJOR STRUCTURES AND EQUIPMENT APRANCEMEffr Refer to the Site Plot Plan, Figure 1.2-1, and the Plant Plot Plan, Figure , 1.2-2, for the site general layout. The plant structures arrangement plans and sections are shown cn Figures 1.2-3 through 1.2-22. - The Turbine Building is oriented parallel to ' State Po.ad A1A and the shoreline l of the Atlantic Ocean, with the Reactor Pulldine located on the east, j' or seaward, side of the Turbine Building. The, Peactor ' Auxiliary Puilding is located perpendicular to and east of the Turbine Building, oriented in an j east-west direction. Tbv Fuel .Pandling Pulldine is located east of the 1 .
, .+
4 1.2-12
- g.
. _ . , . ,, , _ _ _ _ ~ . . - - ~ __ __ _ . _ _ _ . - _ _ . - __ _ _ _ _ _ _ . _ -
4 Racetor Building end tha Reactor Auxiliary Building, oriented in a north-south dir ction. The Reactor Containment duilding encloses the steel containment. structure, which houses the Nuclear Steam Supply System consisting of the reactor, steam generators, reactor coolant - pumps , pressurizer, and other reactor auxiliaries. .The containment structure is served by a polar bridge crane. The ' Reactor Auxiliary Building houses the waste management facilities , Engineered Safety Features' components, heating and venti 11ating system components, electrical equipment, laboratories, offices , laundry and control room.
- The Fuel Handling Building contains the spent fuel pool and new fuel storage facilities, as well as the cooling equipment for the fuel pool. The fuel is transferred from the Reactor Building to the Fuel Handling Building through the fuel transfer tube.
h Wrbine Building houses the turbine generator, condensers, feedwater heaters, condensate and feedwater pumps, turbine auxiliaries and electrical switchgear assemblies and other electrical distribution systeen which are non-Class 1E. 1.2.4 SHARED SYSTEMS AND INTERCONNECTIONS BENEEN UNIT 1 AND UNIT 2 Common St Lucie site facilities that are shared between St Lucie Units 1 and 2, which were engineered, installed and licensed on the St Lucie Unit i docket (Docket No. 50-335) are only described in this FSAR when found necessary for purposes of
,c lar ity for a particular system and/or component description;. otherwise, an appropriate reference to the St Lucie Unit 1 docket is made. The systems and components which are shared (one system which may be used by either or both units) l between St Lucie Units 1 and 2 are discussed in Subsection 3.1.5.
The following is a list of systems interconnected *(one complete system on each unit a which say, under certain conditions, be used by the other unit) between St Lucie } Units 1 and 2: a.) condensate storage tanka, b)' Liquid Waste Management System,
- c) Instrument Air Systes, d) Station Air System, 3
- e) Diesel Generator Fuel Oil Storage and Transfer System, and i
f) startup transformers. A tie between the two units has been provided from the Unit 2 condensate storage j tank to the Unit 1 auxiliary feedwater pump's suction for a backup tornado missile , protected water supply. This is normally isolated. The valve line ups on the St ' Lucie Unit 2 tank assures that the minisasa quantity of water required for ' safe shutdown is maintained at all times. i ,
. 0124F 1.2-13 Amendment No. 5, (4/90)
l The Liquid Waste Management System,is interconnected at two non-seismic, non-safety locations by notinally, locked closed valves. One interconnection allows either unit to transfer liquid wastes to the other unit's holdup tanks. The other interconnection allows St Lucie Unit I to transfer liquid
; waste to the St Lucie Unit 2 Aerated Waste Storage Tank.
The Instrument Air System is interconnected between units via automatically controlled valves. The Station Air Systems are interconnected between units, but are isolated via normally locked closed valves. The Diesel Generator Fuel Oil Storage and Transfer System has a seisste Category I interconnecting tie line between St Lucie Units 1 and 2. Seismic ' Category I locked closed isolation valves assure that the tie line is opened only after administrative approval has been obtained. The startup transformers are sized to accommodate the auxiliary loads of the unit under any operating or accident condition. Each set of start up transformers (IA-2A, 15-25) is provided with a manual switching arrangement i
' which permits paralleling 4.16 kV power to St Lucie Units 1 and 2 under ad- I ministrative control. In the event one of the four startup transformers .
has to be removed from service for repair, the 4.16 kV power to both St Lucie Units 1 and 2 is paralleled to facilitate continued operation of both units. A single startup transformer is adequately sised to accommodate the auxiliary loads of both units under accident conditions when aligned as described above. ; If it should be necessary to start a unit with the other unit operating and. I with one startup transformer sligned to supply 4.16 kV power to both units, appropriate operating procedures assure that the startup transformer is not ! overloaded should an accident condition arise. l 1 j 1.2.5 SECURITY PLAN i l
- As discussed in Section 13.6, a cousson site security plan ts provided for i l St Lucie Units 1 and 2. ,
i I 1.2.6 DfERGENCY PLAN l l j As discussed in Section 13.3, a common site ear.rgency plan is provided fer j St Lucie Units 1 and 2. j I 1.2.7 SYMBOLS AND ABBREVIATIONS ON FIGURES l l , Definitions of symbols and abbreviations used throughout the chapters on l !- fluid and electrical systems are shown in detail on Figures 1.2-23 through _ 1.2-33. The auxiliary pumps P&I diagram is shown on Figure 1.2-34. I
, 1.2-14 i .
. i
l Y l D. SIMPLIFIED PLANT SYSTEM DIAGRAMS l f l FIGURES; DRAWINGS; CHARTS; AND USTS FOR PWR's l ' This section contains simplified one-line " training" diagrams ! which are suitable for display to the Executive Team during I a technical briefing,
- s. Remotor Cooient System
~
- b. Chemical and Volume Comrol System (CVCS)
(including Sorio Acid Systemi i I
- c. Emergency Core Cooling System (ECCS) l 1
o High Head (HHil
- j o intermedises HeadilHil j o Low Head (LHil
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, o Recircuiselon Phase A5enment j o Any Other Plant Unique ECCS
- d. Residual Hest.Removel(RNR) System
- e. Conseinment CooEng System
- f. Containment Spray (CS) System g.
}I AuxBery Feedweenr System (AFW)
- h. Main Steam System
- 1. Condensees and Feedweenr Systern
, J. Emergency Electrieel Distribution System: o Switehyard o 4160/480 VAC o DC System (250/128 VOC) o instrument Power System (120 VAC)
- k. Sofety Reisted Component /Servios Cooling Water Systems j l. Reestor Vessel and intemed Drawings
- m. Vessed Water Level instrument Range Diagremo
- n. Neutron Monitoring Systern Range Diagreme
- c. Conteinment Drawings
- p. ECCS Pump Performance Curves o HHf l o lHI o LHI o CS -
o RHR o APW ! q. AEsoellaneous IBustrative Drawings 4
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EXCORE NEUTRON SOURCE RANGE C-moNmnuNasysTun l FISSION CHAANIER 3d
$ m a N WWE RANGE W i 2 X 10.a % i R*
200 % WR Loo SAFETY , EXTEfGED RANGE I y WWE RANGE l e 4 1 16 % 10 8 CPS F SS ON CHARABER 125 % u R s.Fm. 7 cournotennseests C 0% UIC 200 % O UNIT 1 O y e ' <
. m i
* ! E' iCPS .
is 5 CPS O m t mo nn sm RM l N MR WDE RANGE - 2X104 %
. 200 % !
h 1 CPS BF 3PROPORTIONALCOUNTER 105 CPS i t WR LOG SAFETY WWE RANGE y to4 % 200% FISSION CHARABER , Q3 LNGEAR SAFETY &
-^
CONTROL CHANNELS . 0% UIC 200 % i UNIT 2 i
m - _, - - O I 4 4 i 1 1 ( s 0 4 b i Reactor Containment 4 a i 9 e 4 e 9 9 e 6 9 9 4 0
h a 4 hmits ...sh,a _ ,.gu... .e n ,% m an , a _, ,g _,% , 4 e O 1 REACTOR CONTAINMENT BUILDING L=
~
.j'b~
't
- i l '
ll
~.
. V" l! ,
bi > Y
~~ .. '
~
. o/ -
-~oggga '< .
, w ,e . -._
g -
.. qf({
_ov. c -
. ggR R 2__ __y = k* v ~ - -
.99 _.
= ;
ca 0504013 PAGE 185 FIGURE 30 REV.2 FOR TRANGNO USE ONLY _ _ _ _ _ _ . . __.---_ - ---- - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - -
l 1 l l t I l i l l 1 e 1
\ 1 4
( 1 l ECCS Pump Performance Curves 1 1 1
= l e
i e I t l l 4 6 l 1 i l 4 e
/
, NPSH AT(IMP. , / - 20 g
' _ _ g
- 10<nz C1.
~
B - 0 * - 90 u-HEAD
.z.3000
- 80
$2500 z ,
- 70 -
U ,
'E1500 -
60
< F. 5 z ,
$; 1000 -
50 >-- z
~ ~b.
1 p - 40 $ j 0 0
" . c u.
,, .~ ?
30 "' 3hkh =kg 8= BHP AT 1.0 SP. GR. - 300 - 20 3 = E _"gg 2" m
~ 250 g - 10 e OEm< E r-w g gg-< e-< 5 f.
0
=m! !n, 0 100 200 300 400 500 600 700
.0 3 -$,
- GALLONS PER MINUTE E 5 , ,
1 4 a
,8-x 4
Bi 45: o
.600 HP (1.05.G.)-
- 3IX a:
9g - o i x
- 2m m -
80 - m 4 M LA. EFF. g 2 N - o T HEAD g 200 - b 60 - ' 3 x o a z-E
$50 -
i
= -
9 0bs40 - M r r
/ 3 v5 30 -
o g - u.
% *iB NPSH z 10 -
" es i
MEN'E riti= iii m i;f n i
' 20 - I m =
0 z
- 5. me5 41 2c> 10 -
s2 *Ec e' m q
#I E !g" 0 , ,
m;;; *8 0 1000 2000 32 4000 5000 0 "% 5 Z ? GAU.ONS PER MINUTE
l i REVISION NO.: PROCEDURE 777tE:
'4 PAGE:
APPENDIXES / FIGURES / TABLES 2-EOP "99 EMERGENCY OPERATING PROCEDURE ST. LUCIE UNIT 2 PIGURE 2 A
?
SAFETY INJECTION FLOW VS. RCS PRE 88URE _l l l l l l l l ja l I e , 1 mammpawnminianapiced P ~ 5 Flow. ImesmedRowisisesthenthis l
! jj@ . Mh5WWW N
S jg 3
-I R Q -11 I i m R -
5 p R G
- E i S i
jg . 1MTrah W-
,hopsmilon .
l i
- 2 M Trains a'
i
- hopendon
)
l fgf S m jg $ \ N
% i x x
\
0 , '- N 0 15 20 m m i d 60E m 4 ' s e
- m. - _ . .
l PRESSURIZFD WATER REACTOR UNIT-SAINT LUCIE UNTT 2 El-1 1.0 FACILITY STATISTICS l., Nat Naaw/ Unit: Saint Lucie 2
- 2. Docket Number: 50-389
- 3. Owner / Operator: Florida Power & Ught Co. Fl.1
- 4. Other Planta On Site: Saint Lucie Unit !
- 5. Site Imcation/ Address: bectuosan taland in Saint !acie County beween Fort Nrce Fl.1 and Stuart. 12 miles southesas of Fort Nrce, FL. .
- 6. MAX Facility Ucense71memal 2700 l Fl.1 ,
j Power Umit: (MW:)
- 7. Main Turbine Generator Rated Output: (MWe) 890 *
/ Fl.1
- 8. Ultimate Heat Sink - Supply to Main Atlantic Ocean F9.2.7.2 Turbine Condenser.
- 9. Ultisante Hest Sink - Supply to ECCS Atlantic Ocesa - F9.2.7.2 Service Water SyeWs):
- 10. PWR Type: 2-toop F9.2.1.2
- 11. NSSS Vendor: Consbustan Engmeenag Fl.1
- 12. Turbine Generator W f== F1.2.3.4
- 13. Arch. Engineenag Ebemo services Inc.
- 14. Constructor: Florida Power & Ught Phaann Service Fl.4
- 15. Dete of Operenas Lacenselaamane: 06/10/83 NRC Information Digest
- 16. Date of Commercial Operation: 08/08/83 NRC Information Digest NOTES:
l 4 I e i NRC Plant Infonnation Book 03/93
PRESSURIZED . WATER REACm]: UNIT-SADR - LUCIE UNIT 2 E2-1 l 2.0: REACTOR COOLANT SYSTEM ' t
. 1. MAX TJ ItCS Ps.asme seisty ~ 27s0 T32.1.2 tumie (peis):
- 2. Nusshar of Fuel ^^ 217 TBsJ.I
- 3. Nussber of Fuel Rede per AssemMy 236 TBsJ.I Passeertess- (PER) Level
- 4. PER Hasest Cuandrimel seipous: , ,/ g -g/ 2# d 3 *
-j
- 5. T.S. MIN Oposumeg PZR Imel: 275 T33.43
("Jains) Pressuriser Pow Operated Reuer Valves (PORV)
- s. vah. w a - we.: v.ien pr es.. 4. z, . g fv.1475
- 7. o, - w.3 , . - ,# = se - 77 S. ,oRv s.n.r , tG m ms.u ini>> ~ s.s.
=>
- 9. 70av Power supply tes : le - -
p[ p##IM / 3 von t 1:s 12s msss
- 1 I
L pH p u,a,w \ p guk fr>MIVt s'Yh'+tII bf* ps fi g - fety's sc , 4 y n 't s e 7 < I n,s. n a n e, - A?Acb&hr5) c F N * "" "."~.a "i " '
- j. i NRC Plant Inforunnalen Book
- . . 03/93
]
- ~. . - - ,
PRESSURIZED WNIER REACTUC UNTTSAINT LUCIE UNTT 2 E2-2 Pnesertaer PORY. Overpressure Protecties (NDTT) Setpoint
- 10. OPEN Sapoissa(s) :
JAf .
- e. Heasup : (psis) $470(see 'r13.4.9.3 . .
, %A) 6 C**"**" ' ') mI~ ,45J -
aswoo
<<r nf *ik- .. - -
II. OPEN Temperenne 8seposas(s)
.. Ms g, , ( ,) ,
43n
=*=
y g p ps/# aF"N -
- b. CooWoon (deg F)
I i is M$steh8
- 12. No. FORva seguised by TJ.: 2 Ts3.4.9J
- (int NDIT pseesseien wish so asher . ePalmas evadaMe)
- 13. TJ. MIN veas Opennes : (eg.is.) 3.53 Ts3.4.9J (PORv NDIT.hsenserve)
- 14. Osbar T.S. Opesome for NDTT 4esher1hme P90Weaendeems opemage) si,,.c.t. s c~l: , r.,cn a 14;eFS / [$sh. [ENN 1e'.
FORY Block Vahes (if applicable) .
- 15. valve wh No. v 147s y.1477 Ps5.14e
... v., ,ew ,,,, - , n., ut u esy $ f 7 -b010 7*
v.m ,(vm> qqo / Ss6- 4l'2-'] i l Pnesurher Code Safety Vahes { 4 . 4 17. Number of Saeny valves : 3 PFs5.1-4 l Is. TJ. Fiessee assef sapoins :(pain) 2300*1 s 733.4.2.2 4 )I
. )
. NRC Plant Infansmalen Book-03/93 .
,. _ __ .. . . _ . . . . . _ . _ . . _ _ . _ _ - _ , _ _ . , _ _ _ _ . . . _ _ _ . . _ _ . - _ . . _ . . _ _ . ~ ._ _ 1 e , PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 2 E2'-3 Pnesuriser Spray Valves . 5 19.' Velve Ensincesion No. : Il00E !!00F FI%5.14
- 20. RCs pusip ID thes supplies sprey: 232 281- FF35.13 2L. Auxiliary Pressunser sprey supply CVCs F9.3.4.1.1
. sy men:
l l Randor Coolant Pianps (RCP) a
- 22. - Pu p idesm6esties No.: 2Al 2A2 231 282 FFg5.13
}
- 23. Pu.p rower supply mus : 2A 1 2s 1 2s 1 2A-1 FFg5.1-3 aus Vohess :(VAC) 6.9KV 6.9KV 6.9KV 6.9KV FFgt.3-le t.
Reactor Coolant Pump (RCP) Motor Cooling
- 24. syne. A No.e
- c-m _^ Cooling Weser syne. FFs9.2.2.2 4
1 I'
- 25. syne. s m
- u. nem sys eme,.redi o ..am se*=de.)
i c)CICS) e w- Coolant Pianp (RCP) Seal LSection , y y g )( p t d [F N 1 L 27. sy.e. A No.e : / MI' $ F5 I .3 g pe, , 2s.-syne. s m .e: V
- 29. non syne.e me,. red o .. am me=de.)
- Reactor Coolant Pianp (RCP) Thennel Barrier Cooling i
'~ so. sy.e. w m .e: C- r---. ~i-p weser sy.s. F9.2.2.2 54U p 7 jsM pi 3i: sy e. s m.s.
- 32. Both synes.e Required :
dM l 4 he j df <.
- 0.e.. as redundaat) J3 $1"g;4 $ g, i
/ YhrM NM
"*8:
- 1. Two shutdown ecoling solief valve whk a lit oneing of less eben or equel to 350 pois less then M for 16 or has.p l _ducedmuhar$889 (sourse Ts3.4.9.3).
- ' 2.
To.peresure. The values are en follows: Tech spes li.e diSerent esspouse for relist prosecoce based o for b <295 F- e ~ ~ n r ., qoor
- p,4mp pois >4easa ,e.d r s n(rs a .v.6., a m.2a
./ 35opois sor.seid.w.esd p <4we, j6f~J # TT / Th L W. T 7N M *d
- > n,or e. des.,=. = w s i.t.s a q 1sm .7,ru1.,.y)
. NRC Plant Infonmation Book 43/93
1 PRESSURIZED WATE3 REACH)2 UNIT-SAINT LUCIE UNIT 2 E3-1
. 3.0
.ESF ACTUATION TECHNICAL SPECIFICATION SETPO 4
l 1
\
I
\
l i i l . 1 I i i i h i 4 ) l NRC Pinot Infennaden M - 03/93
TABLF 2.2-1
; .4 ~ REACTOR PROTECTIVE INSTRLMENTATION TRIP SETPOINT LIMITS r-FUNCTIONAL UNIT 8 TRIP SETPOINT ALLOWABLE VALUES -
- 1. Manual Reactor Trip Not Applicable Not Applicable g- 2. Variable Power Level - High "'
'4 , m Four Reactor Coolant Pumps s' 9.61% above THERMAL POWER,
?
Operating with a minimum setpoint of s 9.61% above THERMAL POWER, and a minimum setpoint of 15% of ' 15% of RATED THERMAL POWER, a ximum of s 107.0% of RATED THERMAL POWER and a maximum and 'a m' of s 107.0% of RATED THERMAL POWER. RATED 1HERNAL POWER. t
'3. Pres'surizer Pressure - High s 2370 psia s 2374 psia
- 4. Thermal Margin / Low Pressure"'
l ,
' Four Reactor Coolant Pumps Trip set point adjusted to not Operating Trip setpoint adjusted to not exceed tne limit lines of exceed the limit lines of I" Figu'res 2.2-3 and 2.2-4. Figures 2.2-3 and 2.2-4.
[ . Minimum value of 1900 psia. Minimum value of 1900. psia.
- 5. Containment Pressure - High s 3.0 psig s 3.1 psig t
- 6. Steam Generator Pressure - Low a 626.0 psia (2) 2 621.0 psia (2)
- 7. Steam Generator Pressure"' s 120.0 psid s 132.0 psid
- { ,
Difference - High (Logic in TM/LP Trip Unit)- ; t I ! T*
' 8. Steam Generator Level - Low 2 20 5% (3) i m 19.5% (3) = ! * ~
e - 3
- Ii I i
. r i.
p _ m . _ _ . - m_ .__ - _ _ _ _ _ _ _ _ _ . _ _ . - m
C- ( ( ( TABLE 2.2-1 (Continued) U ' REACTOR PROTECTIVE INSTRislEldTATION TRIP SETPOINT LIMITS - fUNCTI0lML UNIT TRIP SETPOINT i r- ALLOWABLE VALUES ! 8 9. Local Power Density - High
- Trip setpoint adjusted to !
m Operating Trip setpoint adjusted to '
. not exceed the limit lines not exceed the~ limit lines c of Figures 2.2-1 and 2.2-2 .of Figures 2.2-1 and 2.2-2.
$ 10. Loss of Component Cooling Water m'636 gpm**
a 636 gpa p i to Reactor Coolant. Pumps-Low ,
- 11. Reactor Protection System Logic Not Applicable Not Applicable 1
- 12. Reactor Trip Breakers Not Applicable Not Applicable
+
- 13. Rate of Change of Power - High" s 2.49 decades per minute s 2.'49 decades per minute
[
- 14. ' Reactor Coolant Flow - Low"' = 95.4% of design Reactor a 94.9% of design Reactor l Coolant flow with four Coolant flow with four 1
pumpsoperating* pumps operating
- I
- 15. Loss of. Lead (Turbine) = 800 psig 1
. Hydraulic Fluid Pressure - Low
- a 800 psig ~
- u. I I
1
- Design reactor coolant flow with four pumps operating is 363,000 gpa.
1
**10-minute time delay after relay actuatioq. . ;
t 3 ' c. E
, l O
- - . . _ . _ . . . . . _ . . . _ _ _ _ . . _ _ m. . . . _ _ ___
!" ,y 1AutE 2.2-I '(Continued)
- r. REACIOR PROTECIIVE INSIRilM[NIATION IRIP SEIPOINT LIMIIS
.n.
m-o TABIE NOTATION g (1) Trip may be manually bypassed below 0.5% of RAIED TIEltMAL POWER during testing pursuant to 5gutclal
'q. Test Exceptiod 3.10.3; bypass shall be automatically removed when the IlERMAt- POWER is greater - - "
m than or equal to 0.5% of RAIED TIERMAL POWER. ~ (2) Trip may be manually bypassed below 705 psig; bypass shall be automatically removed at or above 705 psig.
~
(3) % of the narrow range steam generator level indication. . (4) Trip may be bypassed below 10 *% and above 15% of RATED 1HERNAL POWER; bypass shall be automatically. ' removed when llERMAL POWER is > 10 4% or 5,15% of RAIED ilEllMAL POWER. (5) Trip may be bypassed below 15% of RATED TIERMAL POWER; bypass shall be automatically removed when
~
sy lilERMAL POWER is greater than or equal to 15% of RA1ED IlERMAL POWER. l I l . i W i
+ .
_ _ . . _ _ _ _ _ _ _ _ _ __.. ____1._ -._.a. _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _
- v , -
{. - . - - . . . . - . .. __u. l l t 4 w. 1.0 _L_.. ...
.e9.... . . .. . .e.u.
._ .p .,
.- M.. um.. .
=...e.
.. ._m
- 4. . . . . . . .m
_. ..._.+ __ _... . __ &e. _. . . .
.. .. .= mm.
.u... . .
_. s -M..._ _4
. . mi._ m ms. $__ . _
e
- i. . ...e., .
l 9._. . i . .e 0.8 , , , , . , , , , . , , , , , , , , . . , . ,
- -____. . *e u.
sm .
.. .mmm....
- * . . ....4.-i.mmm.m.
.mm 4
.. .. 6
~__..__ .t_....__. . . . , . . ,
. .........s....._
_. .f . . . ;. , . . . . . ... _ . . . . . . .
.~...
. . . . . ...t .- .
. . . ...t __. _...__ -._. _ _....... . .
. . .... . .. . . .._._...___= - #__
. /_.. . .
0.8 . _ _ . .y...__. _- _.=- .. .:. . _ ... . . _. .. .. __ ;- _...g........
. .... . t. .
. _..-...._._._..n_-.,__.
_. y. . ., OR2 ._ . ..-. .
.__s._ . v . .. ,. ... _ ..
. . ._.t.............._.....-
3
. . . = t: . . ......_ .
g.u. _. . _
. q. :.::=.:.q-/-t-- -- r : S. _.LO.P. _.E = 1.0.
0.4 ._. .. . ._.. . . . . . . . . _ - _ . ...t. .. . _........4 , __.___w..._.._...__
..____.__.____.._a._
. . . _ . . .. -. ._.___.t __
L
.. ._ ..._..._ f... =_. .
._._.i__._...,.._.._.._..~- .. __. . . . _ ..._..._.. _._. ...... _ ....
0.2 . _ . _. . .
. ....._. _.. p. ,_._. . . ,_ ....,_. .._.,. _ . ..
. - - - - _.2._..
,. . . ... . i...__._ _ . _ . . .
... . _. _. _.....t ......_....
_ -t._._ _.. _... _ _ ... . . . -
. - ~ _ _. ._.__.._.... -
. .. .= . . . . . _ ... . . _ . _ . . .
l. 0'0 O.0 0.2 0.4 0.6 0.8 .1.0 FRAC. TION OF RATED THERM AL POWER 1 Figure 2.21 I Local power density . High trip setpoint Part 1 (Fraction of RATED THERMAL POWER versus QR2 I
. 1 I
t
)
I l er ,,,,........ ,- e i
1 #
- 1. 4 UNACCEPTABLE *
; UNACCEPTABLE '- OPERATION 1.2 - OPERATION i -
_4 C 1* 0 - (-0.22,1.0 (0.22. 1.0) t -
- 0. 0 -
~
QR 2 ~ 1
- 0. 6 . - ' '
l
~ '
(-0.60,0.50) (0.60,0.50) ' ~ j 0.4 - }
~
! ~ ACCEPTABLE OPERATION 0.2 - 4
; 0. 0
~
i . , , . , , , , , , i
-0.6 -0.4 -0.2 0.0 0. 2 0. 4 0. 6 4
AXIAL SHAPE INDEX (Yg ) ' FIGURE 2.2-2 . LOCAL POWER DEMI!TY,-HIGH TRIP SETPOINT PART 2 (QR2 versus Y g ) m i ST. 'LUCIE - UNIT 2 2-8 Amendment No. 42
I i-1
~
.i
- i
- i 1
- 1.70 1 4.
1
.. . . . . . . - . . . . _ - . . 1 1 60 WHERF A x ORi = 0u d- _ __._.
i aun otMP unn . n L7.85 x.T 9410 var 'una n
* - ' ' - - - ~ ~
1.50 - l
\a ._n ninet i i ,,
1.40 - \1 .
\; . .
A
\
i 1 -
\ . ..
1.30
\ '
- \ /
. \ /- _ .
' \ i '
c 4 rw a v . ww . r,] v.wwi
- '1.20 -- -
g j p 4
- t
\: - - ' '
- T _fa=mm X
' ' ~ ~ ~ ~ ~ ~
l . '1 10 . i mr, we , 4
. . . ~ _ __ __._.. ._ _.
. 1.00 * -._ _
~
j- -0.6 -0.4 -0.2 0.0 0.2 0.4 d .'6 i AXIAL SHAPE INDEX, Y) , l '
- 3. +
4 l 1 I i l l ! FIGURE 2.2~'J THERMAL MARGIN /LCW ?RESSURE TRIP SETPOINT PART 1-(Y j ersus V A)) ST. LUCIE-UNIT 2 2 Amendment 30. 8 .
9 WHERE: Aj x QR) = QDNB AND Ptrip var = 1400 x QDNB + 17.85 x Tin - 9410 . 1.2 - - r 1.0 0.95 O'.95
/J ~ ~ ~ ' ~ ~
0*8 / __.:_._ a s0 .. . l
/ -
i i /
- / -
QR 0.6 -- -- 1 / - . / ~D' '
/
0.4- / . _ . - .
)
0.2 __ _ . _ . _ _ _ _ . 1 b 0.15 4 0.0 -- - 0.0 0.2 0.4 0.6 0.8 1.0 1.2 FRACTION OF RATEQ THERMAL POWER d FIGURE 2.2-4 THERMAL MARGIN / LOW PRESSURE' TRIP SETPOIllT PART 2 (FRACTION OF RATED THERMAL POWER VERSUS QR)) ST. LUCIE - UNIT 2 2-10 Amendment No. 8
. "TA8lE 3.3-4 d
y E ENGINEERED SAFETY FEATURES AC10A110N SYSTEM 'INSIRtetENTATION TRIP VAltES
, FUNCTIONAL UltiT Att0WABLE TRIP SEIP0lijf VAttES C
E 1. ,5AFETY INJECTION (SIAS) H a. Manual (Irly Suttons) Not AppItcable Not Applicable
- b. Containment Pressure - High 5 3 5 p519 $ 3.6 psig l
- c. Pressurizer Pressure - Low 1 1736 psia 1 1728 psia
-d. Automatic Actuation Logic Not Applicable Not Applicable .
l
- 2. CONTAllGENT SPRAY (CSAS)
- a. Manual (Trip Buttons) Not Applicable Not Applicable sa -
g b. Containment Pressure -- High-High 5 5.40 psig 5 5 50 psi 9 l cc Automatic Actuation Logic Not' Applicable Not Applicable I
- 3. CONTAIISIENT ISOLATION (CIAS)
- a. Manual CIAS (Irlp Buttons) Not Applicabie Not Applicable
- b. Safety Injection (SIAS) ~Not Applicable Not Applicable i
- c. Containment Pressure - High 5 3.5 psig 5 3 6 psig l
- d. Conteinment Radiation - High < 10 R/hr i 10 R/hr '
S e. Automatic Actuation Logic Not Applicat'- Not Applicable z . l P 4. MAIN STEAM LINE ISOLATION
, a. Manual (Trip Buttons) Not Applicable Not Applicable
- b. Steam Generator Pressure - tow > 600 psia-l
> 567 psia
,' c. Containment Pressure - High 5'3.5 psig 1 3.6 psig l g d. Automatic Actuation Logic Not Applicable Hot Applicable
4 o - TA8tE 3.3-4 (Continued) Y . y ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES E ALLOWA8LE
, FUNCTIONAL LWili IRIP VAlHE VALUES
.k 5. CONTAINMENT St#GP RECIRCULATION (RAS)
)
- a. Manual RAS (Trip Buttons) Not AppIIcable Not Applicable
- b. Refueling Water Storage Tank - Low 5.67 feet above
~
4.62 feet to 6.24 feet tank bottom above tank bottom
- c. Automatic Actuation Logic Not Applicable Not Applicable
- 6. LOSS OF POWER
- a. (1) 4.16 kV Emergency Bus Undervoltage (Loss of Voltage) 1 3120 volts 1 3120 volts y (2) 480 V Emergency Bus undervoltage
* (Loss of Voltage) 1 360 volts > 360 volts
[ b. (1) 4.16 kV Emergency Bus Undervoltage 1 3848 volts en (Degraded Voltage) 1 3848 volts with a 10-second with a 10-second time delay time delay (2) 480 V Emergency Bus Undervoltage
. .(Degraded Voltage) 1 432 volts 1 432 volts
- 7. AUXILIARY FEEDWATER (AFAS) g a. Manual (Trip Buttons) Not Appilcable Not Appilcable
- b. Automatic Actuation Logic Not Applicable Not Applicable
, c. SG 2A828 Level Low > 19.0%
c,
> 18.0%
,E 8. AUXILIARY FEED!!ATER ISOLATI0ll O a. Steam Generator aP-ligh 1 275 psid 89.2 to 281 psid
- b. Feedwater Header AP-liigh 1 150.3 psid 56.0 to 157.5 psid
_ _ - _ . _ - _ _ _ _ _ - _ _ _ .- _ _ _m__ _ _ _ - _ _ _ _ _ _ - - _ _ _ ____-_ _ _ _ _ _ _ __ _ _ - - _ _ _ _ _ _ - -. _ _ _ _ _ - _ _ _ . - _ _ _ - _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _
, PRESSURIZED WATER REACTOR UNIT SAINT LUCIE UNIT 2 E4-1 l4.0 CHEMICAL AND VOLUME CONTROL SYSTEM (CVCS)
- 1. . Sysassa Naas : CVCS P9.3.4.1.1 CYCS Pumps
- 2. Pump basiseesses No.: 2A 28 2C FFs9.3-Sc
- 3. Pum , Type: PDP PDP PDF F1b9.34 (FDP er Cesariengel)
- 4. Pump Power Supply Bus : A 2, d2@ M 2Ad !I.*M.
Bue Vohees : (VAC) Y$0 yy0 @ " p
- 5. e) Pump Design Flow Rese:(spea) 44 44 44 fib 9.34 b) Reissed Diesherge M-(pe0 I Mgt( %g~r [ s t 6.
- WIY Q ,
uah Cherries /Mekeep Speism Plow Rete : (spa) 132 FTb9.3-7
}$$[/
g g ,,
/
j ( q j (Euse of MAX Chersing Plow and Nemiset Seal Weser %)
- 7. Pusup shesed for High Head (Safety) No No No P9.3.4.1.1 34eemos :
S. Pumpe sustion esces eaed to Spees Puel Pool? W3 hN 3,17-si,mf S " f021, 7 hN l f i l Pump Coellag Water Systems
- (IsshMie Labo Oil Coolers, Seal Coches, Moeor Cooling, Roose Coohes, see)
- 9e 1) Cooler Name
- M 8k
- 2) Syenem *A* Nases:
l
. 3) Syenom *B* Name:
- 4) Beeb Syenome Requered 0*'EE I i
9b !) Comier Name: i'
- 2) Synem 'A' Name:
- 3) Syees *3* Name: ,
- 4) Bos Syenome h 0.e.. gg seduasleet) i s NRC Plant Information Book 03/93
.......-__..-.m. . _ _ ~ _ . . _ ._ . _ _ . .
{- PRESSURIZED WATTR REACTD2 UNITSAINT LUCIE UNIT 2 E4-2 i. 9e !) Cooier Naass:
; 2) System 'A' Nems:
1 i i 3) System 'B' Nems: i'
- 4) Besh Sysenes Requesed: 0.e.. als redendom) ,
9d I) Cooler Nees: i 1
- 2) Sysase 'A' Name:
a
- 3) Sysase "B' Name:
- i 4) Rudi Syename Regn ed. 0.e.. als sedundam) j i'
i' Beste Acid Piemsps F-- .___7 Beration) 1
- 10. he , Names n sie 4and us,t., n.,
l ,
- 11. hap h h - No.: 2A 25 FFs9.3-56 J
j 121 hay rype : , com com me.3-sb (FDP er CapriaginD
- 13. h e ,ro rSuppay h e: Mee 2d6 p,cc ggg
[dMM .. v
- as.vobe.e (vac) t/se t/* 8 t
! .i4. a) n., % ri Reas:w to le N.3-20 0 t:b i b) Reissed Deseherpe Head:(ps0 Ib Ib ! .. t.w . Nas I) Nas 1) [A s
- 15. hop ih.
or* ,lI404 76 L . I wM ' y j M !. 8.rie Acs4 s.,,u M17-my ) i6. re.t A' h 2A h p(ib [C/d /N/Cf */, % FFs91Sb l
- n. re.m : Nees: 2 k
- u v 1
2 Nam, x - @.p.tef~n .e n - 4::: -ypovatos _ m .
+ -nte ng 4 ... --
i I NRC Flant Worumaties Book g3y e
PRESSURIZED WNIER REACTOR UNIT. SAINT LUCIE UNIT 2 E6-1 i 6.0 l INTERMEDIATE HEAD (SAFETY) INJECTION SYSTEM (IHI) j j i 1. Symem Name High Pressure Safety Imiection Symem . i i i 2. Pump (e) erosseed between Meat Unies Nees - FFs6S1 - 1 l . I
; 3. IHI Pump Susesse Crieerie : I of 2 P6.3.2.2J I
'wh a= Number of Pumpe)
. ) i ] 'IHI Pumps
- 4. Pump le==asAcence No. :
2A ff/15 23 ff/ 3".C FFs6SI
- 5. Puey Power Supply Bus 2A.3 25 3 'P83.1.1.I(b) {
k he Vohese : (VAC) 41 4. F5J.l.l.l(b)
- 6. e) Pump gaglas Flow Rese:(sym) 345 (see 345 Fn6SI i j
geaW Nees I) l b) ne:med Dnehenge ased:(,si) Myy' % f( fM'Wgr'fu3 t
- 7. e) Pump Bansa Flowmees:(spa) 685 / 685 /
{ FD611 b) Releend Discharge Head:(pel) 356 356
- s. Pem,shmeoerHead (pe') 1231 / FF3 6M
{- Pump Cooling Water Systemas
- l Gestude Labe oil Cooiers. Seal coolsas, hdonor coolsas. Roman Conhes. ess) I 9e 1) Cocier Name: M $ / Q [f[ Pp.2.2.2
- 2) System 'A' Nesse: C l Coohes yueses
\
i 3) Symem 'B' Nesse:
- 4) Bodi Syseene Regesed: G.s.,ggseduedem) a 9b !) Cooler Name:
- 2) Symeas 'A' Nemes:
- 3) Symean 'B' Neens:
i .
- 4) modi syneem magered: 0.e.. ael reduades) i 1
NRC Fisst Intensation boa 03/93 - j i l
PRESSURIZED WATE3 REACIUR UNIT SAINT LUCIE UNrr 2 E6-2 9e 1) Coolet Name:
- 2) Syses *A* Name:
)-
- 3) Syssee 'B' Nasse:
i, _.
- 4) hash Systems Regared. 04.. agg redundaat) i ,
, 9d Il CooierName: . 4
- 2) Syenem 'A' Name:
- 3) Syenom 's' Name:
- 4) Bash Symesse Requesed. 0.e.. gig whendent)
- 1. Iseludse 30 syn bypass now.
4 1 T i b Y l' I 4 F
+
9 e 0 6 NRC Meat Informadoe ' Book M/93
- -. . - - - _ . ~ . . . . . . . _ . ~ - - - - - - _ _ - . - - . . _ - . . _ - . - - . _ . - . -
PRESSURIZED WATT,0 REACIDR UNIT SADR LUCIE UNIT 2 E9-1 l_ 9.0 LOW HEAD (SAFETY) INJECTION SYSTEM i (LHl)/ RESIDUAL HEAT REMOVAL SYSTEM i I. Byese Naas (LIG 6sassion) tow Preesmo safety Isisesion Sysism F6.3.2.2.3 i
- 2. Sysasst Name : OtMR ibastion) Shandows Cooling synem F5.4.7.1.1 5
j General 5 l
- 3. IJt! psovide ausaies for IDG N/A
! Ptampo dwing IDCA Resinculseios l j } Phans ?
- 4. LIG psovies auseine for DG No F%6S2 i Ptsmpe dueing IACA Resissedamos
{ Phase? j Se. , MAX RER h Tempeessee 325 PS.4.7.1.1 . l (desF) 4 M l 56. MAX RHE Design Temposseme:(deg 350 l FTW.3-1 j , P) ' + l 6e. MAX RER Onsranas Pteemme:(psia) 276 PS.4.7.1.1
- M FD6S1 i 6b. MAX RER Denism Pteamme: (peis) 500 y ..
E 6c. RHR Suence IJan Soissy Velve gp ./ F 3.2.6 3 l Relief sapous: (peis) l
- 7. IJG provide eheranas a==h spesy sepabday:
a) IJG synem have ise own aprey No FF36 3-le header? b) IJG symem deseherges isso the - Yes FF36 Ste causeamment Speny System Spesy i Hender ? { c) IJG sysse discharge isso No FF3 6Ste C==- Speny Pese Sonstion ?
- 8. Puey(s) esses tied haswesa Plant Unha No FF36 Sla f .
- 9. IJG (ispection tsaceson) Puerp . I of 2 F6.3.2J.2 suesses Crherie :
Mh number of puespe) -
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNTT 2 . E9-2 LHI/RHR Pumps '
- 10. Nap idenu6cesion No. .
2A b/ff 2BL,/Jf FFg6.3 le -
- 11. hmp Power Supply Bus : 2A-3 28-3 F8..l.l.l(by Bue Vohese (VAC) 4.16KV 4.16KV F8.3.1.1.l(b) ,
- 12. Pump Shusoff Head : (pei) g. .""",'.; L - [ M - " % ii[ ~ h Pisap Cooling Water Systems (Include Labe Oil Coolers, Seal Cooling Motor Cooling Room Cooling, see) 13e 1) Cooier Nees: *""-_*r_, ,
pg 8844+- : 7-,
^
- 2) Syenem *A* Name: .
.-.=i -
illa s
- 3) System 'B' Name:
- 4) Seek Systems Requued 0 *" g redundant) 13b !) Cooler Nees:
- 2) System 'A' Neans:
- 3) Syenem 'B' Name:
1
- 4) Bash Systems Requwed 0 *" E )
l 13e 1) CocierName:
- )I
- 2) System 'A' Name: l
, 3) System 'B' Name: i 4) Both Systems Requesd 6.e., g saiundant) i 13d I) Cooler Neans: i
- 2) System 'A' Nanse
- 3) System 'B' Name:
- 4) Bauh Systems Regeved 0 ** SIU # )
Heat Exchanger hey Side Cooung Water System i 14. Syeese 'A' Name :
- c ;- __ Coohng Weser System F9.222
- 15. Synom 'B' Name : .
- 16. Beek Syeneas Required :
- 6. . 55 eeduadant)
N&tES:
. l 1
,' NRC Plant leformention Book g3/93 !
1
. . . , . . . - . . . . . . - . _ . . . - _ . . ~ . _ , _ . . . . . . . . - . . . - - - - . . - . . _ . - ..
PRESSURIZED WATER REACID2 UNIT-SAINT LUCIE UNIT 2 - E10-1 I i i 10.0 SAFETY INJECTION TANKS L i i. j l. Tank Nems : Sdesy leiection TarA l l i 4
! I i
i 2. Nussber of Tanks : 4 FF36 3-lo
- 3. TJ. MIN Weser Voimans : (mains) 1420A8 '133.5-1
- 4. TJ. MIN Niempos Piessnes. g T53.5-1 4
- F- > ;
.. LP* 9 1 .< . ,
I I i
- i. 1 i
1 l i i I ' e W G I J u i
- 4 f
i I s l i t i 4 e n 3 1 5
t 1 PRESSURIZED WATE2 REACTOR UNITSAINT LUCIE . UNIT 2 E11-1 l 11.0 BORATED WATER SOURCES for HHi,lHI, LHi, CVCS l i l' Rahsellag Water Storage Tank (RWST)
- 1. Teak Nees : Refueling Weser Tank
- 2. ' Task Ideseiaseman No. : 6g FF362 41
! 3s. TJ. nGN eepenity for C : 417,000 / 113.1.2.8 needeag, gag (gen E
i 3b. Igg [T.S. HEN esposity for W (ge0 ', 4.' Design empeosy :(ge0 554,000 / M J.2.2.4 j
- 5. Uameable Tank Volume
- (gen 60.000 M.3.2.2.4 1
- 6. RWrr Supplien CVCB ? Yes
- FF36.2-41
\
i
- 7. RWBT Supplies HIG ? N/A
- 8. RWBT Supplies IMI ? Yes M J.2.2.4 j 9. RW5T Supplies IJG ? Yes M.3.2.2.4
- 10. Tank esessend to Other Unit's Tank ? No j -
FF36.241
- 11. hen RWTT Level for ECCS Swap 5 T3Tb3.34 toe- Suny : (usies)
None 1)
?
)
l (If automano swapever, use TJ. nGN ' j j RW5T 1me0 i i Y % l l l Concentrated Borie' Acid Storage Tasks (BAST) for Eanergency l Beration
- . 12. Teak Name
- Boric Acid niekm, Teak f: 13. Task ha*h No. : 2A 25 FFs9.3-5b 1
- 14. nGN No. of Teaks sequered by TJ. I of 2 T53.1.2.8
- for Opensaans ndode
15e TJ. nGN espeoity for C _ see Nees 2 see Noas 2 h6ade asung :(ge0 , sa Varishie Tsr s3.11 j 15b IglBl TJ. nGN espesisy for gpf opmations : (gen
- 16. Design espeeky
- (gen.
9975 / 9975 / m9.34 (' 17. RAFT have gavky feed to leposeson Yes Yes T53.1.2.2 , . pump asseion? 4 y a. , + - - - wma _ v -+- , -
. ~ - . . - - - . - - . . . . - - . - - . . - . - . . - . - - . - . . - . .----.
- 1 I
1 i
!ITJRE 3.:.-:. S". EIE 2 MIN M" weommunacmma j
i i
}
t [ i : . i
. /I ..
]/ ] ly__.
^
[l ! ! ,/ : g i 8650) j ACCEPTABLE l
/j [
vyyp y r y
}' ;
' i
'g r (7530) l /
) I .~ '
;N[
gg l # I
' [![
]d !
l [665C ) ! i t 08 ' ' -
/ ;
j I l ' V ' h l g i ' L I g l(53!,0-) j 4 5 '- i ..l l' I i g ; j j 1
, j
- l. UNACCEPTABLE j
- . h i !, OPERATION ! i l I l .
i - kl T i l l l N!. , I
^ _ L- l 1 t -
.\ .
3 u u u- 3 M M i (4i96 PPM) (4546 PPM) (4895 PPM) (5245 PPM) (5595 PPM) (5944 PPM) (6294 PP0 i
~ man im cose-( w x hrk e )
~ .
l
- ST.LUCIE - UNIT 2 -
.3/4 i-15 ,
Amendment No. 40
- NY O N & f fil'I
\
PRESSURI7Fn WATER REACTOR UNIT-SAINT LUCIE UNrr 2 E11-2
- 18. Other T.S. Sourcas for Emergency Boretion (other than BAST or RWST) 1([#M[
j
. 2) l NOTES: le . 2 9 N, $nowIfny'3gQsy(,g,pg I Above sank boaom. , p (4 o'$ t{,4 L t e '
2- T13.1.2.8 re(we 2 TS 63ure 3.1 1. I i 1 l l il 4 NRC Plant Information Book . 03/93
j FRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 3 E12-1 12.0
! CONDENSATE, FEEDWATER, and CIRCULATING WATER SYSTEMS . !
l
! 1 l
1 s Camdensate Puseps i j i j' !. hany hh No. : 2A 25 2C FF310.1 2a l 1
- 2. hay Power supply Bus : 2A-2 25-2 . 78J.l.l.l(b) s
- ama vam.oo i(vAc) 9. Ilk? -4.dkv 9. the v A/- 2 <*Io7t' W
- 3. Menor abased udsk (.cadeemsse Boomer N/A j h=,?
- 4. F .e , C. ,eeey
- ( 5 F.n A , ,e, O g e p e r - 3Il his,)
55*/. i
- 3. - No. h ,e - e., !
. aluer Main hadweier hay ope einem ,
- 6. Pump MHead : (psi) oo 0
- A9" 4
l'andamente Beester bps j 7, hamp hh No. : N/A i 8. Pump Power supply Bus : j nus Vohnse :(VAC)
- 9. Pump Capeeky 3 (5 Fuu Poeper per j Pusip)
- 10. Maminumi No. Pumps seqinued for j single Mais Pendwer hug opwesion l : .
, 11. ham, abuso# Head (pei) 1 Mais Feedwater Pussps
- 12. Puey M% No. : 2A 2B PFs10.1-26
- 13. hump Power Sousee : (Motor or museer moeor Pfbl0.41 Besen) l 14. hamp Eleotne Power sugyly Bus: 2Al 251 PB.3.1.1.l(s) nes vahase :(VAC) 6.9KV 6.9KV PB.3.1.1.l(s)
- 15. Pump Capeeky (5 Pun Power per p Pump)
NRC Plant Interunsties Beek e3/93
MIESSURIZE3 . WATER REACTOR UNIT-SAINT LUCIE ' UNIT 2 E12-2 i 4 Other Feedwater Isdection Pumpe
- 16. Syanes Name N/A
! 17. Pump idemai6eemen No. :
- 18. Pussy Power Sowoe : (Electric or en===)
.l
- 19. Pump aiseine Power supply mus: .
i aus whee. :(VAc)
- 20. e) Pump Qadga Flow Rans:(sps )
i b) malened Die herge Head:(pel) . i 4 5 . Ceedesser Ciraalsting Water Sysimm I i i
- 21. Utnesse Been sink : l Asianus oness P9.2.1.2 ;
d' \ l ) 5 I' l e 9 8 4 4 1
. NRC Flaat Wormation Book gy
4 i
. PRESSURIZED WATER REACTOR UNTT-SANT 1.UCIE IMTT 2 E13-1 i 13.0 AUXILIARY FEEDWATER WATER SUPPLY SOURCES 1
1.0 Water Source #1 J a. Soutes Name: Coh Storess Tank Unit 2 4
- b. Souses hardsmed to AFW Puge Yes FFs!0.12 4
l' Sussion ? 4-
] e .' MIN TJ. Volume (unies) //h
- d. MIN T.S. Volume (se0 307.000 TS3.7.1.7 i
- e. D iga c.peeny : (s O 400.00 Pn9.2-Il
- f. Tank reallable by : (Name)
- D'eser Systemp[ ^g 7,_ ; 1,I
,. 1) F9.2.8.2 j 2) Ch Syseem Unit 2 f/hCluA/IM FFg10.1-2e W sin- ~;e > H ssiv , M-1)
SL_c ' '% : , X-_.-
, , I i ,
I g.
-~ w- "
, , f __; f_) - .m d
j . i 2.0 -Water Source #2
- a. Souseo Name
- ch Storage Tank Unit I .
)-
- b. Sousee had gand to APW Pumpe Yes
, Suomea? 4
- e. MIN TJ. Voluene : (uains) h
- d. MIN T.S. Volume (ge0 116,000 TS3.7.1.3 f
- e. ' Design capacity.: (se0 250.o g ,
g,, g n N.216y, ,,jeg fy; A w-<3'
.r f.
4 Taak ev611eble by : (Mame) 1)
"l- .,- ;___ 4 Q,, Unit i FF310.12c
- 2) Ch System Unig 1 [/NC/ V'b Unit 1 FF310.12c g e n o p h ~ a r s r v M r)
- 3) r ' Wassf Syssen Unit 2 Unit 2 FFs10.12a 4;
1
/
! 4) Coadseness Unit 1 Unit 2 FF310.1-2a l~ i , I 7.
- b. s d
d . .
. Y,
- NRC Plant Infonmation Book- , '03/93
ParAtURIZED WATER REACTUG UNTT-SAINT LUCIE UNTT 2 E13-2 3A Water Soaree 03
- e. . Sowse Nees:
((
- 6. e. w i.uw %
Sustice 7
- e. MIN TJ. Volume : (uniss)
- d. MIN TJ, Vokame : (gel)
- e. Design Capacity : (3s0
- f. Tank mallable by (Nems) 1) .
2) 3) 4) 4A _ Water Searce #4
- s. Sense Name:
ff
- 6. sommehad:sisedto AFWhope aussion f
- e. MIN TJ. Volume : (usias)
- d. MIN TJ. Volume : (sel)
( j e. Design Capacity : (ge0
- f. Tank mallable by (Nees)
{ 1) 2) i 3) i i i. 6 e e
. . . - . _ . - . . . - . . - - - . - . .-. . - - . - . . - ~ _ . . - . . . . _ . . . . - . . . . . . -
l l
- i. . PRESSURIZED WATER REACTOR UNTTSAINT LUCIE ' UNIT 2 E13-3 I
i 5.0. Water Source #5 I- s. Souse Names: M , b. somos bestsang to APW Pumps
- , sussion ?
- e. MDt TJ. Voisms : (uaise)
- d. h0N TJ. Values (sel)
- s. - Design Capeoisy 2 (s O j t. Tank meishis by n (Name) )
- 2) .
- 3) .
. 4) . 6.0 Water Seeree #6
- e. Samme Neess:
fh
- b. -- m w '
sustien ?
- s. MDI TJ. Vohame : (unles)
I l d. htDi TJ. Vohness : (gel) I l
- e. nedse cepeeky :(s o I
- f. Tank melable by
- (Name) 1) i1 4 3) l l
4) l 4 m 4 ,
- NRC Not Inforunation Book 03/93
, FUESSURIZED WATER REACTOR ' UNIT.SADfr LUCIE UNIT 2 E14-1 14.0- AUXILIARY FEEDWATER SYSTEM
- l. Synden Name: Auxiliary Feedwater symem Syeese Canagurados
- 2. Ptamp h h No.: 2A 25 2C FF310.12b
- 3. assem Osammeer ID Ser su possible M M FFg10.1 2b W lia"P' 8 24, A8 M,18 24Ad
- 4. Piney Amoimisines t Yes Yes Yes F10.4.9J S. Pump QonMind besween Piana daies : No No No FF310.1-2b i-l Assikery Feedwater Puenpo
- 6. Pissy Mh No. : 2A 25 2C FFgl0.1-2b
- 7. Puey Type : egenartlessme/Diess0 M M , 2_ _
FF310.12b
- 8. Pump Einsens Puoer supply Bus: 2A3 233 FE.3.1.1.100 Bus Vohese :(VAC) 4.16KV 4.16KV ' gth % C- PS.3.1.1.l(b)
- 9. se os ser to mm. avian man, f:.;,;., FF310.1.le
$4}b
- 10. MIN Design Steams Pseamme for
, 50 FFB30.41
'hmbias creamon :(reis)
!1. a) Fump Engsa Flow Race:(spec 300 / 300 / $10 #
RIhlfLA* - b) m.inted u.ed (peo yyy WM %% s.
- 12. ladisses au somees of wease supply so #1, F2 #1, #2 #1,#2 FF310.1-2b eneb pump massion :
(Weest Souse Nesmber tous Sect.13) Puesp Ceoung Water Systeens '
.0mstude lashe 00 Cooism, Seal Cooling, Menor Conhes. Room Coohes, see) 13e 1) Cooler Name: $loe/( 1
- 2) System 'A' Nesse:
- 3) Sysse 'B' Nams:
- 4) Bedi syness 3.que.g 0.e., nel redundnes)
NRC Plant latermeden - seek . e3/m ! l
.:__--_-----_===----------=-------------
_- - ~-
PRESSURIZED WATED REACIT)R UNIT SAINT LUCIE UNIT 2 E14-2 13b !) Cecaer Name:
- 2) Sysases 'A' Naass:
, 3) Symem 'B' Nems:
- 4) Souh Syeses Seguised. 04 . alg sedundem) 13e 1) Cooist Name:
- 2) Synsas 'A' Nees:
- 3) Synsa *3' Nees:
- 4) nah Syness maquimd, 04. agg sedundsm) 13d I) Canist Nees:
- 2) Symem 'A' Nesse:
- 3) Sysism '3' Nems:
- 4) Bash Syness Requis.d. 04 . als adund**) .
f ' l ' l L l 9 2 1 1 t t i 4
I PRESSURIZED WNIED REACTOR UNTT-SAINT LUCIE UNIT 2- E15-1 15.0 MAIN STEAM SYSTEM
- l. Nueser of essess Osassesare : 2 FF310.1.le
- 2. s s. ,aseet :
3910 M cg yacle < fI f f%Vd
- 3. Tasal Husser of Mais sanssa times :
pan doenic . 2 FF3 10.I.la
- 4. Total Nunser of Mais sesem leoiseson 2 FF31 0.I la.
Valves : F10J.2
- 5. uh TJ. Main assem code 1000 Tsn4.74 Seisey Vdve senpaus : (peig)
- 6. Seema Osasseaur Blowdown Sysassa Yes F10.4.8 availeWef
- 7. Webias typene Vdves Total cepestry ' 45 5 / F10J.3
- (5 Fuu Steam I.aed)
- 8. ?^ Duay Vdves Total M- #
ce,es y , <5 F.M w 4 pur 295 - - t 9 t _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ - _ - _ _ _ _ - - _ - _ _ - _ - - -
_._ - . . _ _ _ , . . _ . _ - . ._ . _ _ . . _ . _ _ . _ . _ _ . . _ . . ~ . . - . . . _ _ _ _ _ . _ _ . _ . . _ PRESSURIZED WATER REACIDR UNIT SAINT .LUCIE UNIT 2 E16-1
. .16.0 REACTOR CONTAINMENT i
+
- 1. CasammmmesType* Large Dry 5.
- 2. MIN TJ. h- Pues Volume : 2.5 108 T15.2.1 - '
(se A) , 1
- 3. Tesel Zisonesues levegeery in Cose s ' 59,00s FTb4.210 l . M
- 4. Design Teenpenmus 1Jesi : (e ) r 164 T35.2.2 ~ !
i ! 5.' Design Imammi Pteemme 1Jeals: (peig) a4 T35.2.2 ' 6e TJ. MAX 1ack Raan : J.505 T33J.1.2 ? (5 h erjM per 24 hee) I l Gb TJ MAX 1Ask Rete Pesempet (peig) GUI - OU $ $ 7334.1.2/- P I "
, f f' '
Hydragas 8====h8===s i 7. Number er sh : ' 2 M.2J.2.2 , l ' i
- s. mm imennes 7 yu M.2.5.2.2 !
L _
- 9. Cesemismess beve W f @ *
/
- 10. Hydsages Pwee sysseen insannest Ice M.2J.2J
- 11. Hydeses Pwee Shased ? ya M.2J.2J e 'l
. I 4
~ .- - - - . . - - . . - _ . . . . . . - - - . - - - .. -. ..- - - . - . . . . - - . - . - . . - .
PRESSURIZED WATE3 REACTO3 UNITSAINT LUCIE UNIT 2 E17-1 l 17.0 REACTOR CONTAINMENT COOLING SYSTEM
.,n
- l. Syese Nams: Comasemmat Cooling tuan synssa N.2.2.2.2 V '
syssen Coma surassen
- 2. Tossi Num6er of Coolmas Unies : 4 M.2.2.2.2
- 3. e- Cookes Syeses need for Yes N.2.2 ch Dayseammiassans during 14CA er DBA aseident seedseines ?
XJia- Do gg compises this session. Seceses Criteria (Camemimmamme Depreses'instion)
- 4. Masianna Nussbar of Fans p.r i M.2.2.2.2 Coeling Unit: - - // 6 ( (
- 5. Mimmasa Num6er of Cooler Unies f g4 [//f wish Na Comaimmaat samos :
- 6. h Nesser of Coolse Ushe 2 a(4 M.2.2.3 widi i Camasament Sara, Tmin -
h8 C ner vanu , 7 .f ;- ! m/ m;V Untr/. mapo l
,s. U* us is .io. No. :
l 2nv44 2nv4. 2nvdC 2nvdD w.u
"- " * *"'r*= "/1.9 I
aus Vohese : (VAC) , Afk#
'1) r (f & U s3 YS' O
_ts , ([T U h/ 1"[#'/ i
^
l , UNIT-E UNIT-F UNIT 4 UNIT-5 i
,s. Um u- No. : g) pp. p h. gA j Sb. Fem Elsetnc Power Supply mus:
l
- f aus vehese :(vAC) j e bj;W,twf l
j f NRC Flant Information B.ek , t3/y3 i
PRESSURIZED WATER REACTT)R UNIT SAINT LUCIE UNIT 2 . E17-2 Cook Units 'Coollag Water System )
- 9. o sym 'A' Nams: Componse Cooling War Symem g g,2 g 2 b) symem '3 Nems: gpg c) Bah Symeme h. 0.s. 315 M*")
wrza. , l
' \
f N' h f ~+ f .ff 'M ' 31r ~ W r&- '* n'g;'lf)fWg. 1 1 I NRC Maat infenanden Bd 03/93
l PRESSURIZED WATER REACTD3 UNTTSAINT LUCIE UNIT 2 E18-1
\
l l 18.0 l CONTAINMENT SPRAY SYSTEM 1 . t ) 1. Syeen Name: Cominammens Spray Synse F6.2.2.2.1 l !~ j Syulema Coseguradom 2
- 2. ceassimment apsey desiemed sur Yes M.2.2
-i
! esammans ' . -f '
- 3. aMR Mest Rashangen shased wida yes M.2.2.2
{ Comamment Spsuy ? , 4., can RaR symem pamde ensene sesseenment symy . espebilley, 1 i
-N i s. IBRRR system have its own No FF36.2-41 t symy header?
c __ i
- b. ImRNR syese discharges imo 'FFs6.2 41 '
l dec- spey synes spar ==*-' s $ r j o. ImRuR syens desharge hoo No FFs6.241
- commmmunspeyPum, j ause ? '
j 5. Pump (s) cnmme besumen Fine 'Jains No FFs6.2 41 i
$900006 CI$1M(3 (Pantalemament M Er$ESdeS)
- 6. w- Number of F.mps widi Hg
- - c is.e , E* gfA .
j 7. mannesidsman med ommash sons , i c as i oNI.v, t - N_ ., F.mp. - ,, 4- n raaf-/
; comammam cenism :
9 i 1 l
_ _ _ _ ~ . _ . _ . . _ . . _ _ _ . __ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ . . _ _ . _ _ _ _ _ _ _ _ _ - _ . . . _ . . _ _ i l PRESSURIZED WATER REACTT)R UNIT SAINT ' LUCIE UNrr 2 E18-2 C=*=l=-* Spray Pumps 1 FUMP1 FUMF-5 PUMPC PUMF-D . So. Pg IdessiSeeuos No. I 2A 2B (( ((, FF36.2-41 9e. Pusip Power Supply Bus: 2A3 233 4 Pl.3.1.1.l(b) Bus Vohnse : (VAC) 4.16KV 4.16KV F8.3.1.1 l(b) Ile Piumpe used for seassiement epsey via Yes Yes P6.2.2.2.t s Suser seelseulseinst 12e Fempo Imested RfIEE Consumment: No No FF362-41 4
. c.,g^ a: ' - FUMF-E FUMF-F FUMFG FUMF-H l.
4 Ob. Pump Idemissasion No. : h[ (( M gg . !, 9b. Pump Power Supply Bus: j - - - - . Bue Vohess : (VAC) I i
- lib Pumpe used for sonnemumens apsey via i Suesp seemeuission?
I
- 12b Pumpe1menced M Comesimmunes.
i s Penny Coeung Water Systemas amenuds take oil cooless, snel conhas. Meier coolies, maass coniing, see) t 13e 1) Cooler Name: A/O sv(
- 2) Symems 'A' Nesse:
- 3) System *S* Name: 1 l
l
- 4) Book sysamme Regumed 0.e.. as admadas) f 13b I) CoolerName:
- 2) sysaem 'A' Name:
- 3) Syssem 'B' Name:
- 4) Bodi Syeesaw Roguesed. 0e. 38 M * )
13e 1) Cooler Name: '
- 2) Synese 'A' Name:
- 3) sysemes 'B' Name: .
- 4) Basi Syssesse Requised. O * . Elg mdualem)
S
- NRC Plant Infonmation Book .
03/93
I
, PRESSURIZED WATED REACTOR UNIT-SAINT LUCIE UNIT 2 E18-3 '
134 1) Coolst Nems: .
- 2) Sysses 'A' Name:
- 3) Synom *3' Name:
1 l
' 4y som syn.m.m.g n.4 0
- M " duad **) I
< l Hong FwAnager Grandary Side Coeung Water Systeet .
j I4 *1*** *A' N=
- Wfco m p o u . f c..),' y fyghn, 15.* Synom *8* Name s s f f'
- i . 3- m m , o .. . 6 e >
t NOTES: 1 l t 4 s. i ' a e i l i )
, 'l-
\
1
- i
- l i' 1 l
1 4 1 1 I I , a 9 I
\
j i. U e
, _ -- - -- ,. . - - . . . . . . - . _ . . .- ~. . . . . . . - . . - ..~ . . . . _ . . .
~
PREF-SURIZED ' WATED ' REACTOR: UNIT SAINT LUCIE UNIT 2 E19-1 19.0: ALTERNATE AC POWER SOURCES Class IE Einvgency Diesel Generator (EDG)
- 1. Tonel Number of EDO 04 : 4 FFs8.3 le J
- 2. Number of EDO for this Plast unis : 2 (See Nees FFg8.3-1 I) i l
3.' Number et EDO shesed between this 0MnFFg83-3
- rism Unisand eihar unkst 4
l<mt.' 3gs .' . DO-A DO-B DOC DO-D I 4e. EDO h No. : 2A 25 * [h Fnt.31 je, r,iesel h%s) : GM-EMD GM-EMD Fnt.3-1 y ; Vt - gg m
., 4 g a
- 46. EDO "M No. : fYh* fk hh' fh St. Dismal u%) :
- 6. No. of Disesi Fast seemse Tasks : 2 Pn9.5-1 1
- 7. Capeshy of seamos Temke :(se0 43,430 Fh93-1 i i )
Class 1 E Diesel Coellag Water Systemme g w. , l 1 j to I) Caeler Names selfCooled fliph [4al/4O p F9JJ.2
~
- 2) sysism 'A' Nees
- 3) synne 's' Name:
l 4) Bah sysasens W. 04.. alg adumdess) 4 ll Sb I) CoolerName: i' l j 2) symem 'A' Name: - l. ' I
- 3) symem 's' Name:
- 4) Besh Symmes Required. Od M I I \
l 8e 1) Cooler Nems: I
- 2) Systsat 'A' Name:
- 3) Synes 'B' Nesse: '
, .- . 0. . - >
i i 4
PRESSURIZED WA'IT3 REAC"IOR - UNIT. SAINT LUCIE UNIT 2 E19-2 Bd I) Cooler Nees: . I
- 2) Syene *A* Nees: l
- 3) Symne 's' Nems:
- 4) amh Sysses Regund 6.e., an adualem) '
Other Clas is Eanergemey Power source. ]
- 9. Souse Names :
- 10. Samme ww M..
- 11. Elsemiset has ID Time . AM Possibis Dinos fies ,
Other ca is s rs cy r r searcos c uas weser sysseam 12e 1) CoolerMeans:
- 2) Syene *A* Nees:
'3) Speam *3* Nesses i
- 4) Bah Syenes Requesd. Od. ERI Ndsedm8) 12b I) ComierNees:
- 2) System 'A' Names:
- 3) Synsas *B* Nasms:
- 4) Bodi Systems Regumed. Od.. as sedundam) i-i 12c I) ConorNees:
1
- 2) Synesen 'A' Nesse
i {.. 3) Syness*3* Nees:
.1
- 4 4) Bash Systems Requiseg 0 s 3gsedundes) ]
, 12d I) comist Nems: ,
- 2) Synsas 'A' Namns:
- 3) System *B' Nees 4
l
- 4) Bodi Symeans Requised 04 alg salundess) l ,
i 1 e 1 1 4
.' PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNrf 2 E19-3 l
t Non-Safety Grade Alternate Power Supplies
)
13, Source Name :
- 14. Source ldmai6 cation No. :
15.' Electncel Bus ID Tim All Pouible Direct Ties :
- 16. How laterconneced :
1 NOTES: yry
- 1. Two engese in tendem, with generator in the middle. Value listed is i of geeseosors. There are a total of 4 d- or unit 2 (Sousee FTb8.31).
he 1~b e thit I J ,1 if/&o V rysk Cn k< cres t f.e4 ;f y e e ess<ry by cw/ec% swhy b,.cyr W$ + ud ' fl.r . a bus -e.xc ess f96- cpr97 W p }< s f ared 'Y c- re */ pbl krr of ic yxtb #ra +t. H,fr. (pace; ucri esa r..u.u &n) , h
. #( '
jl I 1 i I 4 NRC Plant Infonnation Book - 03/93
PRESSURIZED WATER REACTOR UNIT SAINT LUCIE UNIT 2 E20-1 20.0 COOLING WATER SYSTEM
- i I
L '
- l. Symes Nome : lateke Cooling Water Symem Cp/ *D' F
/
Systen Contigurat10.
- . s 2. Symem Purpose
- Irmake cooling weser es eink for abe closed loop P9.2.1 l SceFoment ey rbine eoeleggweest and sneem
- 3. Symem ehesed between paest Unise 7 %mr I U t
.2 I
- 4. . Symesisefusy Releend ? Yes P9.2.1 d
S. Closed or Opee Syssemi ? Open 1 P9.2.1 1, . { Success Criteria - Each Plant Unit (Accident Coeditions) 4 l 6. M- Nussber of Pumps 'I of 3 P9.2.1J.! t i ! 7. Manummes Number of HeatE- ' . N/A lh 1 d Pianin
, PUMP-A PUMP-R PUMP-C PUMP-D 8e. Puesp Idessih= No. 2A 28 2C ff ppg 9.2-1 9e. Pump Power Supply Bus: 2A3 2B3 2AB P8.3.1.1.1(b)
, Bue Voltage : (VAC) 4.16KV 4.16KV 4.16KV PS.3.1.1.l(b)
PUMP-E PUMP-P PUMP-O PUMP-H 8b. Pump Idensi6cesion No. : d M kh 9b. Pug Power supply Bus: Bue Vohese : (VAC)
. . j e
9
' 4 NRC Plant lafonmation Book . 'g3/93
l PRESSURIZED WNIER REACTOR UNTT-SAINT LUCIE UNIT 2 E20-2 '- Pusap Coollag Water Systeams Gachads Imbe 00 Coolers, Seal Cootag. Masor Coolmg, Room Coohag, me) g 10e 1) Cooler Heess: L,--[ [ f a [j h [ a. N' F9.2.1.2 l
- 2) Symem 'A' Nems: SWCooled MM$ [ h.2.1.2
- 3) Symem *B* Names: P9.2.1.2
. ,, ,c,,, , - Re ,,,,,d 0.e = ~) ! 10b 1) Cooler Ness: . l
- 2) Sysses *A* Name:
. 3) sysaem *B* Nesse:
- 4) amh sysses a*guu*g, 0.e ant reduadao 10o I) Cooler Name:
1
- 2) Synom *A* Nems:- 1 1
l 3) sysses *3* Name:
- 4) Bosh SyeemsItaquased. 0#* ele )
l 10d 1) Comier Nesse: l
- 2) Syness *A* Nasms:
- 3) syseems *3* Naase:
- 4) Both Symeme Requised. 0.e.. als %)
l l l Hegg Fwbanye Secondary Side Coollag Waler Systems i II. Sysases *A* Nesse :
- 12. Syssess *B* Nasme :
- 13. Bosh Systema Regared 0.e., ggs redundant)
NRC Plant Inforunation Book 03/93
, . . _ _ _ _ _ . _ _ _ . _ . ~ . _ . _ _ _ _ . _ .
PRESSURIZED WATES REACTOR UNITSAINT LUCIE UNrr 2 E20-3 WARNING (Generic Statement),
, This matrix below list all cooling water system pumps and associated heat -
exchangers which can be used to maintain, a specific component operable usina the full can=hility of svetam cross-ties andJnterconnections. It sl ould be recognized that under norn al circumstances, the cooling water system may be operated as isolated " trains" to prevent common mode fallures. Consequently, cross-tied alignments of the system may not be procedurally permitted.
- 14. Most Land Matris ID Reformase ,
Maois Pump hAae No Matrix Heat = = ^ __
!D ID Identificenes No A 2A A B 23 3 C 2C C D D E E
.=
F F* G G I
- H H
! 15. Heat Land Masrix i PUMPS HEAT EXCHANGR i COMPONENT NAME & ID i A B C D E F 0 B A B C D E F
- G H e
n Coc.her Weser Heat X X X , Emohenger 2A (Eawee PPs9.2.2) . e- ; Cachsg Weser Heat X X X i t . 23 (sousee FPg9.2.2) i i e e 6 I } xac neer ws n ne aves
PRESSURAZED WATER REACTOR UNTT-SAINT LUCIE UNIT 2 E20-4 l e L f l i 1 L l, i l I i l
- )
i 1 1 i i i E i NOIES: 1 t i . NRC Phmt Infonnation Bock ' 03/93
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNrr 2 ' E21-1 I21.0 COOLING WATER SYSTEM
- 1. symem Nam : , component cooling her sysism P
Systen Configuration
- 2. symen Nepose : Piova best eisk for seassor suailiary symeme under normal P9.2.2.1 opmenes med abuteowna , provide been sink for esteey reissed componesse --* wish rosesor deoey best removal for esis W or DBA eoedition.
- 3. system ebend between plant Unite ? '*--
^"
_.'^ . (( ppg 9.2-2
, 4. symme sessy Boissed ? te. P9.2.2.2 '
- 5. Cleend or Open syneen 1 r u .e y,n Success Criteria - Each Plant Unit (Accident Conditions) 6, u- Number of hope : I of 3 79.2.2J.1
- 7. Maassum Number of Hess Essbenser I of 2 P9.2.2.3.1
- j Ptemps PUMP-A PUMP-B PUMPC PUMP-D 5
, Se. Neo E. Mh No. : 2A 2B 2C g PPs9.2-2 l 9e. Nap Power supply Bus: 2A3 233 l 2AB P8.3.1.1.l(b) 1 Due Vobese : (VAC) 4.16KV 4.16KV 4.16KV i Pl.3.1.1.l(b) 1
- PUMP E PUMP-F PUMPG PUMP-H tb. Pump ideen6eenen No. :
[ hk [k [ 9b. hap Power supply Bus: i Bue Vohege : (VAC) d s
- - NRC Plant Infore==*ian Book s 03/93
- _ - - - - --__- - - . - - . . . .. _ , -... .. - _., . ~ . - . . . . . . - . - . - . .~
> . f, FRESSURIZED WATER REACTOR UNTTcSADrr LUCIE UNIT 2' E21-2
, Peny Cooling Water Systerns ,
(lesnude Lebe Oil Coolere, Seal coolies, Motor Cooling, Room Cooling, etc) (( ' ..- los ' )t ' Cooler Name: .
- 2) System 'A' Name:
- 3) System 'B' Name:
- 4) Boek Sysessee Required: OA. 85 h) 10b - 1) Coolar Name:
- 2) System 'A' Nesse:
- 3) System 'S' Nees: I
- 4) modisymeme w . 0*. am neundnes) 10s 1) CoolerName:
- 2) Sysesse 'A' Name:
- 3) Systems *E' Name: .
- 4) Badi systems Regesired: 0*'E8 )
10d 1) CoolerName: '
- 2) Synese "A" Nases:
1 2
- 3) synem *s' Nases:
- 4) amh symmes seguned. 0 4, as )
i Heat Exchanger f= - f--, Side Cooung Water Systein i 11. System 'A' Name : ' lateke Coches Weser sysse FFs9.21 4
- 12. Sysesse 'B' Noems :
! '13. Bodi Syssene Required : 0.e., Rg redusdem) i !: 0 1 i i e F 4 4 e NRC Flaat Infonastice Book
, .03/93'
' PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UhTT 2 E21-3 WARNING I 1
(Generic Statement) , l This matrix below list all cooling water system pumps and associated heat exchangers which can be used ! to maintain a specific component operable usine the full can=bility of svstem cross-ties and interconnections. I It should be recognized that under normal circumstances, the cooling water system may be operated as isolated ' trains
- to prevent common mode failures. '
i Consequently, cross-tied alignments of the system may not be procedurally permitted. 1 i l l
- 14. Hese imod Meerix ID Refwenos
' Meeria Pump Identiaseman No Meeria Home E ^( '
ID ID IdemnSeamos No A 2A A 2A l s 23 5 1s C 2C C f D D 2 E E j F y J 4 c a
. H H 1
e t
* " Pf*I'rved eek P
- 15. Heat Load Meerix PUMPB HEAT EXCHANGER COMPONENT NAME & ID A B C D E F G H A B C D E F G H Shutdown Heat Exehenger 2A,2B X X X 1 X X C=- Fem Cecier 2HVS 1 A, X X X X X 2HYS-1C
, C=== Fan Cooler 2HVS-1B, X X X X X ) 2HVS-ID , High Presano Seisty leyeonce Puey 2A, X X X 'X X . 2a . a 5. Commel Room NC 3A. 33,3C . X X X i X X Fuel Puesp Heat Exchanger 2A,23 X X X
- a X X
- i. . .
NRC Plant Infm=== dan Book 03/93
+-
__,__ _ . . , . . . - - . - - ~ - - - - - - - - - - - - - ' - - ~ " " ' ' ~ ~ ~ ' ' ' ~ ~ ~ ~ ^ ^~~~~~~ ~ ~ ^
~'l l
PRESSURIZED WATER REACTOR UNIT-SAINT LUCIE UNIT 2 E21-4 RCP Motor Cooling 2A1,2A2,282,281 X X X I X X RCP Thennsl Betrier 2A1,2A2,232, X X X X X 231 Sample Heat Eschenger2A,2B,2C,2D X X X- X X Boric Acid Concentreason2A.2B X X X X X Wome Oes Compressore 2A. 2B X X X X X th Heat Exchaager X X X X X Cesarol Rod Drive Air Coolere X X X X X i i {' s i' l 1 i 4 1 4 i= N O HCS: 4
- 1. Comyceans coolang woest pungw may be aligand to bander A or B.
+
4 6 1 1 j NRC hm Monnados Book
' OM3
_- ~ . . . ... __. , . ~ . . . . . . . . . . - . . - ... - .- . - . - . . - . - 9 PRESSURIZED WATER REACIOR UNrr. SAINT LUCIE UNIT 2 EA-1 l Appendix A- Abbreviations I i AFW Auxiliary Feedwater CVCS Chemical and Volume Control $ystem DBA Design Basis Accident dog Degree (Temperature) I ECCS Emergency Core Cooling System I EDO Emergency Diesel Generator 'I ESF Emergency Safeguards Features ' gym Gallons per Minute ' HEI High Head (Safety) Injection I IHI Intermediate Head (Safety) ! Injection ' LOCA Loss of Coolant Accident LHI Low Head (Safety) Injection MAX Maximus MIN Minimum NDTT Nil Ductility Transition Temperature NRC Nuclear Regulation Commission - NSSS Nuclear Steam Supply System l PDP Positive Displacement Pump ; PORY Puumatic Operated Relief Valve l psi Pounds per Square Inch l PWR Pressurised Water Reactor ' PER Pressuriser RCP Reactor Coolant Pump RCS Reactor Coolant System ' RRR Residual Beat Removal i RWST Refueling Water Storage Tank sq in Square Inch VAC Volts - Alternating Current ) i i Beforence Botations , i F' Final Safety Analysis Report
~
(FSAR) FFg FSAR Figure FTb FSAR Table TS Technical Specifications (TS) i TSFg TS Figure [ TSTb TS Table , TSBasis- TS Basis Section D Licensee's simplified Drawings (Providad in l i Section D of Plant I Information Book) i NRC Flast Intensedem Beek 1-M/M
.]
4g .4 Ng "4. REGULATORY INFORMAfl0N TRACKING SYSTEM. FY93 (BEGINNf NG 1/1/93) - 4 SUM g FAC DOCKET TOTAL , NAME, NO HOURS
' S&W FUEL CO. .07001201 414.0 414.0 BASC0CK & WILCOM Co. 07000027 3300.4 BABCOCK & WILCON Co. 07000824 - 239.0
- 3539.4 BELLEFONTE 1 05000438 461.0 KLLEFONTE 2: ,
05000439 186.3 . 647.3 BROWNS FERRY"1' 05000259 648.0 BROWNS FERRY 2 05000260 5115.1
' BROWNS FERRY 3 05000296 1769.0 7532.1
. BRUNSWICK 1 05000325- 5473.8 BRUNSWICK 2 05000324 8586.2.
14060.0 ,
- CATAWBA 1 05000413 2629.0 CATAWeA 2 05000414 2674.7 5303.7 CRYSTAL RIVER 3 05000302 4216.3 ;
4216.3 FARLEY 1 05000348 2646.5 FARLEY 2' 05000364 2416.0 5062.5 GENERAL EldCTRIC Co. 07001113 1410.7
)
1410.7 i MARRIS 1 05000400 4218.2 4218.2 i 1 RATCH 1 05000321 2421.5 ' NATCH 2 05000366- 1912.0 I l 1 01/10/1997 14:39:46 .
- PAGE 1 i-9
\
{
'ih . +
f9 :L
.4 REGULATORY 'INFORMATION TRACKING SYSTEM FY93 (8EGINNING 1/1/93) i
= SUM FAC - DOCKET ~ TOTAL - ;
NAME NO HOURS
- 4333.5 t
MCGUIRE 1' 05000369 .1905.2 MCGulRE 2 05000370 2128.7
- 4033.9 NORTH ANNA 1 05000338 2874.0 -
NORTH ANNA 2 05000339 1987.2
*. 4861.2 NUCLEAR FUEL SERVICES,'INC. 07000143
....3907.5.....
3907.5 s 1 OCONEE 1. 05000269 2887.9 OCONEE 2- 05000270 2012.0
-OCONEE 3 05000287 1842.1
*. 6742.0 ROSINSON 2 05000261 '4359.7 4359.7 ST LUCIE 1 05000335 2196.2 ST LUCIE 2 05000389 2132.5-
- 4328.7
. i
. SUMMER 05000395 4009.3 I l
'4009.3 SURRY 1 05000280 2363.2 SURRY.2 05000281 2415.3
+
- 4778.5 1
TURKEY POINT 3 05000250 2213.7 TURKEY POINT 4 05000251 2414.5 ) I 1 4628.2 V0GTLE 1 05000424 3235.1 V0GTLE 2 05000425 2848.4 l l l 01/10/1997 14:39:46 i PAGE 2 1 9 e
, 4 .
4 t t
, . .. . 4 ,. ..- - . - . . . - . . .. _ . . _ . ~ . . -
.c. +
.f,.Y
- i_.. . . . ~1
' iqr
- Je - REGULATORY INFORMATION TRACKING SYSTEM-
. FY93 (BEGINNING 1/1/93) ,
,. . SUM .
, FAC' - - DOCKET . TOTAL-NAME~ .NO. ,
HOURS' :
.............................. ........ .......... -3
-6083.7'
. WATTS BAR 1- 05000390 14398.9 WATTS BAR 2'. 05000391 3642.2
- 18041.1
~WESTINGNOUSE ELECTRIC CORP.- - '
07001151 1301.5 4 4
- 1301.5
, assssssssa )
'117813.0 +
s t 1 i f I i 1
- 1 1
l .t-
]
i
- 01/10/1997'14:39:46- ' -
PAGE 3 l I m I e , 0 4 '
, B e 6
4 ;\ t 4 t M'- s '
?
, . . , ~ . _ . - . - , ,
-: r .
.g-4 REGULATORY INFORMATION TRACKING SYSTEM FY94 4
$UM FAC- DOCKE1 TOTAL -
NAME NO HOURS 84W FUEL Co. 07001201 497.0
- 497.0
. BABCOCK & WILCOX Co. 07000027 4690.1 BASC0CK & WILCOX CD. 07000824 280.5
*' 4970.6 BELLEFONTE 1 05000438 225.5 SELLEFONTE 2 05000439 26.5
*- 252.0 BROWNS FERRY 1 05000259 700.0 BROWNS FERRY 2}}