ML20059C455
| ML20059C455 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 08/29/1990 |
| From: | Johari Moore NRC OFFICE OF THE GENERAL COUNSEL (OGC) |
| To: | Anderson G, Bloch P, Johnson E Atomic Safety and Licensing Board Panel |
| References | |
| CON-#390-10785 OLA-5, NUDOCS 9009050064 | |
| Download: ML20059C455 (490) | |
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*t / August 29, 1990' 90530P3:30 3 ! cia itev Peter B. Bloch, Chairman Eliza .h,[Bh; Johns 6d Administrative Judge Admini trat1%l4udge Atomic Safety and Licensing Oak Ridge National Laboratory Board Panel P.O. Box 2008 t U.S. Nuclear Regulatory Commission Bethel Valley Road Washington, DC 20555 Building 3500, Mail Stop 6010 Oak Ridge, Tennessee 37831 Dr. George C. Anderson Administrative Judge I 7719 Ridge Drive, N.E.
Seattle, WA 98115 , In the Matter of FLORIDA POWER & LIGHT COMPANY ' (Turkey Point Plant, Units Nos. 3 and 4) Technical Specifications Replacement Docket Nos. 50-250-OLA-5. 50-251-OLA-5
Dear Administrative Judges:
By letter dated August 28, 1990, the Staff issued the License Amendments for Turkey Point Units 3 and 4, which are the subject of the above captioned proceeding. Copies of the amendments and , I the accompanying staff Safety Evaluation are enclosed for your l information. l I Sincerely, l l LT d21 1 Q F_. /1 6 20 Janica E. Moore l Counsel for NRC Staff
Enclosures:
as stated cc w/ enclosures: Service List > l D p G l
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August 28, 1990 i Docket Nos. 50-250 and 50-251 Mr. J. H. Goldberg Executive Vice President Florida Power and Light Company P.O. Box 14000 Juno Beach, Florida 33408-0420 t
Dear Mr. Goldberg:
SUBJECT:
TURKEY POINT UNITS 3 AND 4 - ISSUANCE OF AMENDMENTS RE: REPLACEMENT OF CURRENT TECHNICAL SPECIFICATIONS WITH REVISED TECHNICAL SPECIFICATIONS (TAC NOS. 63038 and 63039, 55915 and 55916, 55384 and and 55385, 71864 and 71865) The Commission has issued the enclosed Amendment No.137 to Facility Operating License No. DPR-31 and Amendment No.132 to Facility Operating License No. DPR-41 l for the Turkey Point Plant, Units Nos. 3 and 4, respectively. The amendments I consist of changes to the Technical Specifications in response to your appli-t cation transmitted by letter dated June 5,1989, as supplemented July 12, 1989, ! November 3, 1989, February 13, 1990, May 1, 1990, June 21, 1990 and July 20, 1990. l These amendments replace the entire current custom Technical Specifications with a set of revised Technical Specifications (RTS) which are based on the standard Technical Specifications for Westinghouse-designed plants. The RTS , include added surveillances, action statements, and an improved user-friendly ! format 65.well.as other technical improvements which have been developed as a ; result of many years of industry operating experience and analy ns since the ; initial TS were licensed for Turkey Point. The RTS impose many additional i requirements and actions for surveillance, testing, and maintenance of equip-ment, and taken as a whole, are believed to represent a significant safety l improvement for the operation of Turkey Point units 3 and 4. The RTS issued with this amendment also serve to close out, for Turkey Point .) units 3 and 4, a number of generic technical issues which have been of concern to the NRC staff at nuclear plants in the U.S. These include the following:
- 1. Generic letter 84-15 (TS to reduce cold fast test-starts of diesel generators, TAC Nos. 55915 and 55916).
- 2. Generic letter 85-09 (TS for testing of shunt trip reliability, Salem !
ATWS Item 4.3, TAC Nos. 55384 and 55385). j
- 3. Generic letter 87-09 (Applicability of Limiting Conditions for Operation, TAC Nos. 71864 and 71865).
- 4. Letter dated 11-28-78, Schwencer (NRC) to Uhrig (FPL) (TS for contain- '
ment purge). r
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I Mr. J. H. Goldberg August 28, 1990 ! l By this letter, we are closing TAC Nos. 63.038 and 63039, 55915 and 55916, 55384 l ard $5385, and 72864 and 71865. i A copy of the Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice. , Sincerely, 1 Gordon E. Edison, Sr. Project Manager ' ProjectDirectorate11+2 . Division of Reactor Projects - I/II r Office of Nuclear Reactor Regulation
Enclosures:
- 1. Amendment No.137to DPR-31 !
- 2. Amendment No.132to DPR-41
- 3. Safety Evaluation cc w/ enclosures:
See next page I I _ ... -._.--,.y-i __
r e i , It Mr. J. H. Goldberg Florida Power and Light Company Turkey Point Plant ec: , Harold F. Reis, Esquire Administrator Newman and Holtzinger, P.C. Department.of Environmental 1615 L Street, N.W. Regulation Washington, DC 20036 Power Plant Siting Section State of Florida Mr. Jack Shreve 2600 Blair Stone Road Office of the Public Counsel Tallahassee, Florida 32301 Room 4, Holland Building
- Tallahassee, Florida 32304 Regional Administrator, Region II.
U.S. Nuclear Regulatory Commission
. John T. Butler, Esquire 101 Marietta Street, N.W. Suite 2900 Steel, Hector and Davis Atlanta, Georgia 30323 4000 Southeast Financial Center Miami, Florida 33131-2398 Attorney General Department of Legal Affairs Mr. K. N. Harris, Vice President. The Capitol Turkey Point Nuclear Plant- Tallahassee, Florida 32304 Florida Power and Light Company P.O. Box 029100 Plant Manager Miami, Florida 33102 Turkey Point Nuclear Plant Florida Power and Light Company P.O. Box 029100 County Manager of Metropolitan Miami, Florida 33102 Dade County 111 NW 1st Street, 29th Floor Miami, Florida 33128 Senior Resident Inspector Turkey Point Nuclear Generating Station U.S. Nuclear Regulatory Commission Post Office Box 57-1185 Miami, Florida 33257-1185 Mr. Jacob Daniel Nash Office of Radiation Control Department of Health and Rehabilitative Services 1317 Winewood Blvd.
Tallahassee, Florida 32399-0700 Intergovernmental Coordination and Review Office of Planning & Budget { Executive Office of the Governor The Capitol Building i Tallahassee, Flori6d 32301 i
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QRIDAPOWERANDLIGHTCOMPANY DOCKET NO. 50-250 , TURKEY POINT PLANT UNIT NO. 3 . . [ AMENDMENT TO FACILITY OPERATING LICENSE ; Amendment No.137 License No. DPR-31
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Florida Power and Light Company (the licensee) dated June 5, 1989, as supplemented July 12, 1989, i November 3, 1989, February 13, 1990, May 1, 1990, June 21, 1990 and July 20, 1990, complies with the standards and requirements of the l Atomic Energy Act of 1954, as amended (the Act) and the Commission's 1 rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; , C. There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health . and safety of the public, and (ii) that such activities will be i conducted in comf.fance with the Commission's regulations; i
- D. The issuance of this amendment will not be inimical to the common l defense and security or to the health and safety of the public; and l- E. The issuance of this amendment is in accordance with 10 CFR Part -
51 of the Commission's 'egulations and all applicable requirements , have been satisfied. l l 1 i
-- -______________._.----n_--_ _ _ _ _ _ . _ _- , .---o , - - - w , .- ,, 4,
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- 2. Accordingly, the license is amended by changes to the Technical l'
. Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License 3 No. DPR-31 is hereby amended to read as follows: :
(B) Technical Specifications and Environmental Protection Plan ! The Technical Specifications contained in Appendix A,
;; as revised through Amendment No.137, are hereby incorporated in the license - The Environmental l
Protection Plan contained in Appendix B is hereby l incorporated into the license. The licensee shall ! operate the facility in accordance with the Technical , Specifications and the Environmental Protection Plan.
- 3. This licent.e amendment is effective as of the date of issuance. This !
license amendment shall be implemented prior to entering mode 6 for restart l following a complete defueling of the reactor, or within 12 months, which- l ever occurs first. , FOR THE NUCLEAR REGULATORY COMMISSION Herbert N. Berkow, Director ProjectDirectorateII-2 Division of Reactor Projects - I/II Office of Nucleur Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 28, 1990 i L 1 l l I L. t ll
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b[/ - i FLORIDA POWER AND LIGHT COMPANY DOCKET NO. 50-251 1 TURKEY POINT PLANT UNIT NO. 4 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.132 License No. OPR-41
- 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Flor.ida Power and Light Company (the licensee) dated June 5,1989, as supplemented July 12, 1989, November 3, 1989, February 13, 1990, May 1, 1990, June 21, 1990 and July 20, 1990, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules:and regulations set forth in 10 CFR Chapter I; , B. The facility will operate in conformity with the application, i the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; t l' D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and i E. The issuance of this amendment is in accordance with 10 CFR Part L 51 of the Commission's regulations and all applicable requirements L have been satisfied. l I ____-__-_---_____-___-_--__-____l
~ 2. Accordingly, the license is amended by changes to the Technical
. Specifications as indicated in the attachment to this license i amendmet, and paragraph 3.B of Facility Operating License No. DU 41-is hereby amended to read as follows:
(B) Technical Specifications and Environmental Protection Plan ' The Technical Specifications contained in Appendix A, - as revised through Amendment No.132, are hereby incorporated in the license. The Environmental , Protection Plan contained in Appendix B is hereby incorporated into the license. The licensee shall operate'the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
- 3. This licente amendment is effective as of the date of issuance. This !
license amendment shall be implemented prior to entering mode 6 for restart following a complete defue',ing of the reactor, or within 12 months, which-ever occurs first. FOR THE NUCLEAR REGULATORY COMMISSION h ' Her ett N. Berkow, Director l Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical ; Specifications t Date of Issuance: August 28, 1990
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'..f: ) i c-ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 137 FACILITY OPERATING ~ LICENSE NO. DPR-31 b
AMENDMENT NO. 132 - FACILITY OPERATING LICENSE NO. DPR-41
~
DOCKET NOS. 50-250 AND 50-251 1 Revise Appendix A as follows: ' s Remove Pages Insert Pages All All
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.x SECTION; PAGE 1.0 DEFINIT 10NS................................................... 1-0
, 11.l'-ACT10N.......................................................-. 1-1
- 1.2 ACTUATION LOGIC TEST.......-................................... 1-1 ,
1; 3 ANALOG CHANNEL OF ERA 110NAL TEST. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 1.4 AX1AL FLUX DIFFERENCE......................................... 1-1 s
-1.5 CHANNEL CAllBRAT10N........................................... 1-1 1.6 CHANNEL-CHECK................................................. 1-1 1.7J CONTAINMENT INTEGRITY........................................... 1-2 1.8 CO NT RO L L E D t r. A KAG E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2' 1.9 CORE ALTERATION............................................... 1-2 1.10 DOSE EQUIVALENT l-131........................................ . 1-2 1.11 E-AVERAGEDISINTEGRATIONENERGY.............................. 'l- 2 L1.12 FREQUENCY N0TATION........................................... 1-3 11.13 GAS DECAY TANK SYSTEM........................................ =1-3 1.14 IDENTIFIED LEAKAGE........................................... 1-3 1.15 MEMBER (S) 0F THE PUBLIC............. ........................ ~1-3 1.16 0FFSITE DOSE CALCULATION MANUAL..... ........................ 3 1.17 OPERABLE - OPERABILITY.............. ........................ 1-4
,; 1.18 OPERATIONAL MODE - M0DE......................................_ 1-4 1.19 PHYSICS TESTS................................................ 1-4 1.20 PRESSURE B0UNDARY LEAKAGL .................................. 1-4
. J' 1.21 PROCESS CONTROL PR0 GRAM...................................... 1-4 1.22 PURGE - PURGING.............................................. 1 .
- 1. 23 - QUADR ANT POWE R 'T I LT R AT 10. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1.24 RATED THERMAL P0WER.......................................... 1-5 1.25 REP 0hTABLE EVENT............................................. 1-5 11.26 SHUTDOWN MARGIN.............................................. 1-5 1.27 SITE B0VNDARY................................................ 1-5 TURKEY POINT - UNITS 3 & 4 i AMENDMENT N05.137AND 132
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it INDEX1 t i oi DEFINITIONS 0 3 j;, -SECTION - V J 3 1.28. -PAGE
} 1.29 s0tioiriCATiON...................... .....
SOURCE ................ 1-5 CHECK................................ f_ . 1.30'THERMAL STAGGERED TEST 8 ASIS........................... 1 l'^ -1.31 .............. 1-5 i' 1.32 P0WER.................................. .............. 1-6 . 1.33 TRIP ACTUATING DEVICL OPERATIONAL UNIDENTiflED ........... TEST. 1-6 1.34 LEAKAGE........................ ................. UNRESTRICTED 1-6 1.35 AREA........................... ................ 1-6 1.36 VENTILAT]ON EXHAUST TREATMENT SYSTE TABLE 1.1 VENTING............................M......................... 1-6 FREQUENCY 1-6 TABLE 1.2 NOTAT10N.......................... ........... 1-7 OPERATIONAL M0 DES..-......................... ........... 1-8 TURKEY POINT - UNITS 3 & 4 ii AMENDMENT N05.137AND '32
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2 .[, 1. 2 9 50 V R C E L C H E C K . . .. . . , . . . . . . . . . . . .- . . . . . . . . . . . . . . . . .. .. ., . .' . . . . . . . . . . . . . / - 1-5s , F*p
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1-5 O,. :1.31 T H E RMA L P0WE R . .". .. . . . . . . . . . .. . . ,' z. . . .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 , @f* 1.'32 ? TRIP ACTUATING DEVICE l0PERATIONAL: TEST.....................-.. 6 1.33 UNI D E NT I F I E D L E AKAG E . . . . . . . . -. . . . . . . . :. . . . .. . . . . . . . . . . . . .;. . . . >. .. . _ 1-6 ' 4: F" x.. '
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,.N 11'. 35 VENTII.ATION: EXHAUST TREATMENT. SYSTEM; . . ... ... . . . . . . . . . . . . . ... . . . 'l : 1. 3 6 ' V E N T I NG . '.' . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . .- . . . . . . . . . . . . . . . . . . 1'- 6 ' w, TABLE lil F REQUENCY ' NOT ATI ON. . . . . 3 . . . . . . . , . . . :. . . . . . . . . . . . . . . . . . . . 1-7 ,n A TABLE 1.2 OPERATIONAL MODES........ . ................................ 1-8 - ,a, :
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" SAFETY LIMITS AND-LIMITING SAFETY SYSTEM SETTINGS-e ;;i'-
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- < 5 SECTION! .
PAGE.
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@i 1 ' 2.I' SAFETY LIMITS > 4 12.12 1 - R E ACT O R 'C0 R E . . .. . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . .. . . . .. . . . . . . . 12-1
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, 2.1; 2 REACTOR C00LMT- SYSTEM PRESSURE. . . . . .~. . . . . . . . , ..... . . . . . . . . . ...
2-1 1 a ,4 ? , FIGURE 2,1-11 REACTOR--CORE SAFETY-LIMIT - THREE LOOPS IN OPERATION.. 2-2 }
.. , 2. 2 LIMITING SAFETY: SYSTEM SETTINGS- , . . -i+ .
x 2.2;1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS.................. 2-3: yy 4 e LTABLEl 2.2-l' REACTOR TRIP SYSTEM INSTRUMENTATION' TRIP SETPOINTS. . . .
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.p_ q 2.1 SAFETY LIMITS 4 4
2.1.1 REACTOR C0RE............................................-.... ' B 1~- 2.1.2. REACTOR COOLANT SYSTEM PRESSURE............-................. LBL2-2 1 2.2 LIMITING SAFETY SYSTEM' SETTINGS-2,2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS............... B 2-3 -l i I
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i l TURKEY POINT - UNITS 3 & 4 iii AMENDMENT N05137 AND 132
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E LIMITING CONDITIONS-FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE a 3/4.0 APPLICABILITY..........................'....................... 3/4 0-1 s , < 3/4.1 REACTIVITY CONTROL SYSTEMS ., c a 3/4.1.1 B0 RATION CONTROL Shutdown Margin - T Greater Than 200 F.................' 3/4 1 ., avg FIGURE 3.1-1 REQUIRED SHUTDOWN MARGIN VERSUS REACTOR COOLANT BORON CONCENTRATION................................,. 3/4 1-3 Shutdown Margin - T avg Less Than or Equal to.200 F....... 3/4 1-4 Moderator Temperature Coef ficient. . . . . . . . . . . . . . . . . . . . . . . 3/4 1.5' : Minimum. Temperature for Criticality...................... 3/4 1-7' 3/4.1.2 B0 RATION SYSTEMS Flow Path - Shutdown..................................... 3/4'l Flow Paths - Operating.................................., 3/4 1-9
- Charging. Pumps - Operating............................... 3/4 1-11 Borated Water Source - Shutdown........................... 3/4 1-12' 'l Borated Water Sources - Operating........................ 3/4~1-14 4 3/4 1-16 Heat Tracing............................................. } t 3/4.1.3' MOVABLE CONTROL ASSEMBLIES .
Group Height............................................. 3/4 1 1 TABLE.3.1-1 ACCIDENT ANALYSES REQUIRING' REEVALUATION IN'THE EVENT OF AN INOPERABLE FULL-LENGTH R00................ 3/4 1-19 Position Indication Systems - Operating.................. 3/4 1-20 TABLE 4.1-1 R0D POSITION INDICATOR SURVEILLANCE REQUIREMENTS...... 3/4 1-22 Posi tion Indication System - Shutdown. . . . . . . . . . . . . . . . . . . . 3/4 1-23' Rod Drop Time............................................ 3/4 1-24 Shutdown Rod Insertion Limit............................. 3/4 1-25 Control Rod Insertion Limits............................. 3/4 1-26 i FIGURE 3.1-2 R0D BANK INSERTION LIMITS VERSUS THERMAL POWER THREE-LOOP OPERATION............. ................... 3/4 1-27 i TURKEY. POINT - UNITS 3 & 4 iv AMEN 0 MENT N05 137AND 132
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~ 3/4.2.2 HEAT FLUX HOT CHANNEL-FACTOR............................. ~3/4?2-4
, FIGURE 3.2-2 K(Z) - NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT.
9 3/4'2-5 3/4'2-111 3/4 ~. 2. 3 . NL' CLEAR ENTHALPY RISE HOT CHANNEL FACT 0R. . . . . . . . . . . . . . . . .. .t 3/4. 2.~ 4 .Q HDRANT. POWER TIll RATIO................................. 3/4 2-13
-3/4.2.5 Db8? PARAMETERS ...................................... ....
3/4 2-16
'3/h.3 INST (UMENTATION ,
3/4.3;1 REACTOR TRIP SYSTEM' INSTRUMENTATION...................... 3/4:3-l'
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TABLE:3.3 REACTOR, TRIP' SYSTEM INSTRUMENTATION................. 1 3/4 3-2 '
; TABLE 4.3-1. REACTOR TRIP. SYSTEM INSTRUMENTATION SURVEILLANCE.
f: REQUIREMENTS.................................
.. 4.... 3/4 3-8 q L3/4.3.2 ENGINEERED' SAFETY FEATURES ACTUATION SYSTEM-INSTRUMENTATION.........................................., 3/4 3 ' TABLE 3.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM .
1
, INSTRUMENTATION.......................................... 3/4 3-14 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION: SYSTEM
. INSTRUMENTATION TRIP SETP0lNTS...........................
3/4 3-23
' TABLE 4.3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM :
-INSTRUMENTATION SUP.VEILLANCE REQUIREMENTS................ 3/4 3-29
' 3/4.' 3. 3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations................ 3/4;3-35 i
i 1 TURKEY POINT - UNITS 3 & 4 v AMENDMENT N05.137AND 132
s INDEX a o. LIMITING' CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS, I SECTION PAGE l TABLE 3.3-4 RADIATION MONITORING INSTRUMENTATION FOR PLANT 0PERATIONS..................................... 3/4 3-36 TABLE.4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS..................... 3/4.3-39 i Movable Incore Detectors.......;......................... 3/4.3-40 Accident Monitoring Instrumentation....................... -3/4 3-41 TABLE 3.3-5 ACCIDENT MONITORING INSTRUMENTATION................... 2/4 3-42 , TABLE 4.3-4 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE-R E Q U I R E M E NT S ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
. .3/4 3-46 Fire Detection' Instrumentation.........................'.. .3/4 3-47 TABLE 3.3-6 FIRE DETECTI)I INSTRUMENTS FOR ESSENTIAL EQUIPMENT.... :3/4 3-48 Radioactive Liquid Effluent Monitoring Instrumentation... 3/4 3-50
' TABLE 3.3-7. RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-51
- TABLE 4.3-5 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ . 3/4-3-53.
Radioactive Gaseous Effluent-Monitoring Instrumentation.. 3/4 3-54 TABLE 3'.3-8 RADI0 ACTIVE GASE0US EFFLUENT MONITORING } INSTRUMENTATION...........................'............... 3/4 3-55 , TABLE 4.3-6 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE: REQUIREMENTS................ 3/4 3-58' l L i
-TURKEY POINT - UNITS 3 & 4 vi AMENDMENT NOS.137 AND 132
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,, SECTION PAGE. {
3/4.4 = REACTOR' COOLANT SYSTEM: . 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION r Startup and Power 0peration.............................. 3/4 4-1 - Hot Standby.............................................. 3/4 4-2 q
' Hot. Shutdown............................................. 3/4 4-3 a
Cold Shutdown ' Loops Fi11ed.............................
~
3/4 4-5 L Cold Shutdown - Loops Not Filled. . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-6 , 3/4. 4.' 2 SAFETY' VALVES. ; Shutdown.......-........................................ 3/4 4-7 Operating...................,.........................- 3/4 4 i 3/4. 4.~ 3 PRESSURI2ER................... .......................... 3/4 4 ~ 3/4.4.4 RELIEF' VALVES................... ......................... 3/4 4-10 3/4.4.5 STEAM GENERATORS.......................................... 3/4 4-11
. TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION............................. 3/4 4-16 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION. . . . . . . . . . . . . . . . . . . . . . . 3/4 4-17 3/4.' 4. 6 REACTOR COOLANT-SYSTEM LEAKAGE *~
Leakage Detection Systems................................ 3/4 4-18 Operational' Leakage...................-................... 3/4 4-19 j TABLE 3.4-1 REACTOR COOLANT. SYSTEM PRESSURE ISOLATION VALVES...... 3/4 4-22 3/4.4.7 CHEMISTRY................................................ 3/4 4-23 J TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............... 3/4~4-24 TABLE.4.4-3 REACTOR COOLANT SYSTEM CHEMIS RY LIMITS SURVEILLANCE . REQUIREMENTS.................... ........................ 3/4 4-25 3/4.4.8 SPECIFIC ACTIVITY.............. ......................... 3/4 4-26
. FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC .
i,CTIVITY LIMIT'VERSUS i.iRCENT OF RATED THERMAL POWER ! WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1pCi/ gram -i DOSE-EQUIVALENT I-131.................................... 3/4 4-27 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM.................................................. 3/4 4-28 TURKEY POINT - UNITS 3 & 4 vii AMENDMENT N05.137AND 132
j r INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 'PAGE 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System...................................- 3/4 4-30 FIGURE 3.4-2 TURKEY POINT' UNITS 3&4 REACTOR C00LANT' SYSTEM HEATUP LIMITATIONS (60*F/hr)
' APPLICABLE UP TO 20 EFPY.-...................-.......>..... 3/4 4-31
.- FIGURE 3.4-3 . TURKEY POINT UNITS 3&4 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS (100 F/hr)
APPLICABLE UP TO 20 EFPY................................. 3/4 4-32 FIGURE 3.4-4 TURKEY POINT UNITS 3&4 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS (100 F/hr) I APPLICABLE UP TO 20 EFPY................................. 3/4 4-33 TABLE 4.4-5 REACTOR-VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE...................................... 3/4 4-34 Pressurizer.............................................. 3/4 4-35 Overpressure Mitigating Systems........................ . 3/4 4-36 3/4.4.10 . STRUCTURAL INTEGRITY................................. ... 3/4'4-38 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................. 3/4 4-39 3/4.5 EMERGENCY CORE-COOLING SYSTEMS 3/4.i.1 ACCUMULATORS............................................. 3/4 5-1
-3/4.5.2 ECCS SUBSYSTEMS - T avg GREATER THAN OR EQUAL TO 350 F.... 3/4 5-3 FIGUI.E 3.5-1 RHR PUMP CURVE....................................... 3/4 5-6 ll 3/4.5.3 - ECCS SUBSYSTEMS - T avg LESS THAN 350 F...... ............ 3/4 5-9' L.-
l l l- 3/4.5.4 REFUELING WATER STORAGE TANK............................. 3/4 5-10 1 TURKEY POINT - UNITS 3 & 4 viii AMENDMENT N05.137AND 132
.p y 7, .
,U:
f INDEX l
. LIMITING CONDITIONS FOR'0PERATION AND SURVEILLANCE REQUIREMENTS-
=SECTION- PAGE'
-3/4.~6 ' CONTAINMENT SYSTEMS (ATMOSPHERIC TYPE CONTAINMENT) 3/4.6.1 PRIMARY CONTAINMENT f
Containment Integrity.................................... 3/4 6-1 Containment Leakage............................ . .... ..... 3/4 6-2 , Containment Air Locks..................................... 3/4 6-4 , Internal Pressure........................................ 3/4 6-6 1 Air Temperature........................................... 3/4 6-7 Containment Structural Integrity......................... 3/4 6,8 Containment Ventilation System........................... 3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System................................, 3/4 6-12 Emergency Containment Cooling System..................... ^3/4 6-14 3/4.6.3 EMERGENCY CONTAINMENT FILTERING SYSTEM .................. 3/4 6-15 - 3/4.6.4 CONTAINMENT-ISOLATION VALVES.............................. 3/4 6 ; 3/4.6.5 COMBUSTIBLE GAS CONTROL Hydrogen Monitors........................................ 3/4 6-19 3/4.6.6 POST ACCIDENT CONTAINMENT VENT SYSTEM.................... 3/4 6-20 i i TURKEY POINT- - UNITS 3 & 4 ix AMENDMENT N05.137 AND 132
J y- INDEX : r LIMITING CONDITIONS FOR OPERATION AND-SURVEILLANCE REQUIREMENTS - i SECTION PAGE
.f 3/4.7 PLANT SYSTEMS ,
~3 /4.7.1 TURBINE CYCLE ,
Safety Valves............................................ _3/4'7-l' .
- TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY' VALVES DURING ;
THREE LOOP 0PERATION...................................-... 3/4 7 .g , TABLE'3.7-2 STEAM LINE SAFETY VALVES PER L00P..................... '3/4 7-2 Auxiliary Feedwater System................. .............. 3/4 7-3 TABLE 3.7-3 AUXILIARY FEE 0 WATER SYSTEM OPERABILITY................ 3/4 7-5 Condensate Storage Tank................................... -3/4 7-6 Specific Activity............................'............ 3/4 7-8
~ TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE- , , -AND ANALYSIS PR0 GRAM..................................... 3/4 7-9 Main Steam Line Isolation Va1ves......................... '3/4 7-10 Standby Feedwater System................................. .3/4 7 3/4.7.2 COMPONENT COOLING WATER SYSTEM........................... 3/4 7-12 3/4.7.3 INTAKE COOLING WATER SYSTEM.............................. 3/4'7-14' 3/4.7.4- ULTIMATE HEAT SINK....................................... 3/4 7 -'3/4.7;5 CONTROL ROOM EMERGENCY VENTILATION SYSTEM................ 3/4 7-16 l 3/4.7.6 SNUBBER 5................................................. 3/4 7-18 I 3/4.7.7 SEALED SOURCE CONTAMINATION.............................. -3/4 7-22 L
s L l l l L TURKEY POINT - UNITS 3 & 4 x AMENDMENT N05137 AND 132 l l
nm , h a 1 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE: o 3/4.7.8 FIRE SUPPRESSION SYSTEMS Fire' Water Suppij and Distribution System................ 3/4 7-24 Spray and/or Spri nkler Systems. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4-7-27 Fire Hose Stations....................................... 3/4 7-29 TABLE'3.-7-4 FIRE HOSE STATIONS.................................... 3/4 7-30 Fire Hydrants and Hydrant Hose Houses..................... 3/4 7-31 TABLE 3.7 FIRE-HYDRANTS......................................... 3/4-7-32. 3/4.7.9 FIRE RATED ASSEMBLIES..................................... 3/4 7-33 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 AC SOURCES 0perating................................................ 3/418-1 TABLE 4.8-1 DIESEL GENERATOR TEST SCHEDULE......................... 3/4 8-10 Shutdown................................................. 3/4 8-11 3/4.8.2 'DC SOURCES: Operating................................................ .3/4 8-13 TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS..................... 3/4 8-16 Shutdown................................................. 3/4 8-17 3/4.8.3 ONSITE POWER DISTRIBUTION 0perating................................................ 3/4 8-18. TABLE 3.8-1 UNIT 3 ALLOWABLE OUTAGE TIMES FOR UNIT 4 LC/MCC IN0PERABILITY.................................. 3/4 8-21 TABLE 3.8-2 UNIT 4 ALLOWABLE OUTAGE TIMES FOR UNIT-3 LC/MCC IN0PERABILITY.................................. 3/4 8-22 Shutdown................................................. 3/4 8-23 i TURKEY POINT - UNITS 3 & 4 xi AMENDMENT N05.137AND 132
lyhc , . < m
'. 'INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS
+ ,
SECTION PAGE 3/4.9- REFUELING OPERATIONS ' 3/4.9l1' BORON. CONCENTRATION...................................... 3/4 9-1
- 3/4.9.2- INSTRUMENTATION.......................................... 3/4 9-2
-3/4.9.3 DECAY TIME....-.................-.......................... .3/4 9-3 3/4.9.4 CONTAINMENT BUILDING PENETRAT10NS....................l,.. 1 3/4 9-4 3/4.9.5 COMMUNICATIONS........................................... .3/4'9-5 3/4.9.6 MANIP'ULATOR CRANE....-.................................... '3/4 9-6 3/4.9.7 . CRANE TRAVEL - SPENT FUEL ST0 RAGE L AREAS. . . . . . . . . . . . . . . . . . 3/4 9-7 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level...........-...........,.................. 3/4 9-8 Low Water Level.... ..................................... - 3/4 9-9 TURKEY POINT - UNITS 3'& 4 xii AMENDMENT N05.137AND 132
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3/4. 9. 9; ! CONT AINMENT VENTI LATION ' ISOLATION . SYSTEM. . . . . . . . .:. . . . . . . . j
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.- J3/4. 9 i13 ' RADI ATION MONITO RI NG . . . . .. . . . . . . . . . . . . . . . . . .. .-. .-. . . . . . . . . .. . 3/4 9-14 <
h /." wx S E q 3/4. 9.14 ' ?S P E NT F U E L ST0 RAG E . . . . ;. . . . . . . . . . . . . . . . . . . . . . .. ,. .,. . .- . . . . . . . ,.3/4'9-15 j3
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g (TABLE-3.9:1; SPENT'_ FUEL BURNVP REQUIREMENTS FOR STORAGE IN-d,
; ; REGION II'0F THE SPENT FUEL PIT.......................... 3/4-9-16; e, v., >
.I '
3/4.10 SPECIAL TEST EXCEPTIONS- ' ' N ;v _
.r ip 3 '3/4.10.1 SHVlDOWN .
MARGIN...'............................... ......... '3/4 10-1
~
q ! ;3/4;10.2- GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS... 3/4 10-2 m e - 2 3/4.10.3 PHYSICS TESTS..,......................................... ~ 3/4 10-3 ' cg , y, " 3/4.10. 4; ; (Thi s sneci fication number is not used). . . . . . . . . .-. . . . . . . . 3/4 10 :! 1 E -.
'3/4;10.5 POSITION INDICATION SYSTEM - SHUT 00WN.................... -- 3/4 10-5 1 3/4111L: RADI0 ACTIVE EFFLUENTS n
4 3/4.11.1LLIQUID EFFLVENTS. 1., Concentration............................................ 3/4 11 ? TABLE-4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS .I ':; PROGRAM- a/4 11-2 , n .
\
W Dose..................................................... .3/4-11-5 5 Liquid Radwaste Treatment System..... ................... 3/4 11-6 , 'e . '5I: .[ J V 1 TURKEY POINT'- UNITS 3 & 4 xiii AMENDMENT N05.137AND 132 3 I g I b y -
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-LIMITING CONDITIONS FOR 0PERATION AND SURVEILLANCE REQUIREMENTSL ,
$p', L ' :SECTION PAGE 3/4.1h 2; GASE0VS' EFFLUENTS
-. Dose Rate;.................................................
3/4.11-7 I ! TABLE 4'.11-2 RADI0 ACTIVE GASE0US WASTE SAMPLING'AND-ANALYSIS ~ E y , PROGRAM.......;..............................'............ 3/4 11 E ~ D o s e - : N o b l e ~ G a s e s . . . '. . . . . . . . . . . . . . . . . . . . . . . . . : . . . . . . . . . . . 3/4c11-12 Dose - l Iodine-131,--Iodine-133, Tritium, and Radioactive = 't' [i ' Materia 1Jin Particulate Form.'- _- c.... 3/4:11 u im w Gaseous Radwaste Treatment System........................ 3/4:11-143 4 E x p l o s i v e G a s M i x t u re . . . =. . . . . . . . . . . .' . . . . . . . .; . . . . . . . . . . . . . . . 3/4 11-15" a - LGas Decay Tanks.......................................... '3/4*11-16c
}- i
#,' 3/4.11.3- 50 LID'RADI0 ACTIVE WASTES................................... 3/4 11-17 3/4.11.4- TOTAL 00SE.'..r,......... ....................'............ 3/4 11-18 3/4.12 RADIOLOGICAL' ENVIRONMENTAL MONITORING 3/4.12.1' MONITORING PR0 GRAM....................................... '3/4 12-1
. TABLE 3.12-1 ~ RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........ 3/4 12-3 u
$ TABLE 3.12-2 . REPORTING LEVELS FOR-RADI0 ACTIVITY: CONCENTRATIONS
-IN ENVIRONMENTAL: SAMPLES..........-....,.......-....
. ....... 3/4 12 .
TABLE 4.12-1 DETECTION-CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS.................................................: 3/4 12 6" ' 3 /4'.12 i2 L AND ' US E C E NS U S . . . . . . . . . =. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 12-11' 3/4.-12.3- INTERLABORATOR'l COMPARISON PR0 GRAM................-....... 3/4 12-13
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. TURKEY. POINT UNITS 3 & 4 xiv AMENDMENT NOS.137AND'132
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~ BASES k'
SECTION4 PAGE I; 3/4.0 APPLICABILITY...............................................- B 3/4 0 3/4.1- REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2' B0 RATION SYSTEMS..........................................- B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................ B'3/4 1 i i L 3/4.2 POWER DISTRIBUTION LIMITS................................... B 3/4 2-1 3/4,2.1 AXIAL FLUX DIFFERENCE...................................... B 3/4 2-1 ! i FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL P0WER...............................-.............. B 3/4 2-3 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT-CHANNEL FACT 0R.......................... B 3/4 2-4 B 3/4 2-8 ! 3/4.2.4 QUADRANT POWER TILT RATI0................................. 3/4.2.5 DNB PARAMETERS............................................ B 3/4 2-8 1 3/4.3 INSTRUMENTATION 1 3/4.3.1- and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY l FEATURES ACTUATION SYSTEM INSTRUMENTATION................. B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................................ B 3/4 3-3 , I i L i l TURKEY POINT - UNITS 3 & 4 xv AMENDMENT N05.137 AND132 a
.ie i
INDEX BASES PAGE SECTION
. 3/4.4 REAC10R COOLANT- SYSTEM 3/4.4.1 ' REACTOR COOLANT LOOPS AND COOLANT CIRCULATION............. B 3/4 4-1 3/4.4.2 . SAFETY VALVES....................... .................... B 3/4 4-2 3/4.4.3' PRESSURIZER............................................... B 3/4 4-2 3/4.4.4 RELIEF VALVES............................................. B 3/4:4-3 3/4.4.5' STEAM GENERATORS.......................................... -B 3/4 4-3 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................. B 3/4 4-4 3/4.4.7 CHEMISTRY................................................. B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY......................................... B~3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS............................... BL3/4 4-7 TABLE B 3/4;4-1 REACTOR VESSEL TOUGHNESS - UNIT 3. . . . . . . . . . . . . . . . . B 3/4 4-10' TABLE B 3/4.4-2 REACTOR VESSEL TOUGHNESS - UNIT 4................. B 3/4 4-11 3/4. 4.10 ST RUCT URAL INT EGRITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-16 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................. B 3/4 4-17 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................. B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS............................... B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK.............................. B 3/4 5-2 TURKEY POINT - UNITS 3 & 4 xvi AMENDMENT N05137AND 132
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PAGEs Y <-
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,3/4.9 - , REFUELING 0PERATIONS O>
E , 3/43.1;BORONCONCENTRATION..... '
.......... . . . . . . . . . . . . . . . . . . . . . . . . . B'3/4 9-1 ll B-q l
(3/4.9.2- INSTRUMENTATION..................................-......... BJ3/4j9-l' 3/4.9.3 DECAY TIME...........'...............................=....... B 3/4 9-1~ ]a L3/4.9.4n CONTAINMENT BUILDING PENETRATIONS...... .................. . B' '3/419-1~ !! F 3/4.9.5 COMMUNICATIONS......,...................;................... B 3/4 9-1 3/4.9.6_ MANIPULATOR CRANE..............................:............ . B 3/4 9-2; ui 13/4.9.17 ~ - CRANE TRAVEL - SPENT-FUEL STORAGE AREAS....................- B'3/4 9-2 ,
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n 3/4.9.9. CONTAINMENT VENTILATION ISOLATION. SYSTEM....................-
'B13/4.9-2 j
.-3/4.9.10:and-3/4.9.11 WATER LEVEL - REACTOR VESSEL AND
. STORAGE ~P00L.............................................. B 3/4 9-3 3 3/4.9.12 HANDLING 0F SPENTKFUEL CASK .............;................. :\
B 3/4 9-3 b a
, 3/4.9.13 RADIATION M0NITORING....................-..........-......... B 3/4'9-3i y
= .
1
;3/4.'.14 9 SPENT, FUEL STORAGE.. ... .... .................:......;....
B13/4 9-3 4 3/4.10:'SPECIAL TEST EXCEPTIONS. L3/4.10.1 SHUT 00WN' MARGIN............'................................ B 3/4 10-1 t o 3/4.10.2' GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.... B 3/4 10 [ f, J 3/4.,10.3 PHYSICS TESTS............................................. B 3/4 10-1
' ~
'3/4.10.4 (This specification number is not used)................... B 3/4 10-1 ,
SHUTD0WN..................... B 3/4 10-1 3/4.10 5 POSITION INDICATION SYSTEM - . t I a i
\
. TURKEY POINT --UNITS-3 & 4 xix AMENDMENT N05.137 AND 132 ,
4 2
La INDEX + BASES SECTION PAGE 3/4-11 RADIDACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS........................................ .B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS................... ...................
. B 3/4 11 3/4.11.3 SOLIO RADI0 ACTIVE WASTES................................ B 3/4 11-6:
3/4.11.'4 TOTAL 00SE.............................................. B 3/4 11-6 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM,..................................... B 3/4 12-l'
'3/4.12.2 LAND USE CENSUS.......................................... B'3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM...................... B 3/4 12-2 TURKEY POINT - UNITS 3 & 4 xx AMENDMENT N05 137AND 132
y , r v a ;
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y DESIGN FEATURES- 8 ESECTIONJ 'PAGE , I 5.11 SITEJ
, 5.1.1 . EXCLUSION AREA.....-..........................................=
5 ; 5-1 5.1.2 LOW PO PU L AT I ON 2 0N E . . . . . . . .. . . . . . . . . . . .: . . . . . . . . . . .. . . . . . .. . . . . -.
* < 5.1.3J MAP DEFINING UNRESTRICTED' AREAS AND SITE BOUNDARY lFOR: a 7 .RADI0 ACTIVE. GASEOUS AND LIQUID FFFLUENTS.......,.............-.. . 5-11 ;
y.
. -5.2'= CONTAINMENT' 4' 5, 2.1. CO N F IG U R AT I O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . 5 5.2.2. DESIGN-PRESSURE AND TEMPERATURE.................'............ 1
's FIGUREL S.1-l'. SITE' AREA MAP......................................... 5-2'- . l
' FIGUREL-5'.1-2.-PLANT' AREA MAP....................................... 5- 3 -.
m f 5. 3 REACTOR CORE (( ', 15. 3.1 :. FUE L; AS S EMB L I E S . . . . . . . . . . . ~. . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . .. . . 5-4 (
5.3.2 .CONh]L R00. ASSEMBLIES....................................... 4; 4
1 :; e
- 5.'4"~ REACTOR COOLANT-SYSTEM 5.4.'1 -DESIGN PRESSURE AND. TEMPERATURE.'............................ 5-4 5.4.2::v0tuME... ..................................................... 5-4 !
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- 5. 5 METEOROLOGICAL-TOWER LOCATION.................................. J5-4' 4
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'i V , s '5.6.1 CRITICALITY................................................. 5-5' w
p 15.6.2 DRAINAGE.................................................... 5-6 g'e 5.6.3I CAPACITY... - - 5 . at t I 'I S: ' 5. 7 COMPONENT CYCLIC OR TRANSIENT LIMIT........................... 5-6 ' Q. v
, . TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS.................. 5-7
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t-Yi f: 1 xxi
- 4. TURKEY POINT - UNITS 3 & 4 AMENDMENT N05.137AND 132 2 m ff l.f' ,'; q
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g .6.5 REVIEW.AND AUDIT............................................ 6-5
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6 6'.5;1 PLANTJNUCLEAR SAFETY COMMITTEE Function.................................................... 16-5 1
, ~ Composition................................................
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% 6.' 6 : REPORTABLE ^ EVENT ACTION.,.......... . ....... ................ 6-12 e, i, 14 - I 1 1 T
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6.7- SAFETY LIMIT VIOLATION...................,._................. 6-12' i i1 t-H, +, ' 16.8 PROCEDURES-AND PR0 GRAMS....................-,................ 13-
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6.~9': REPORTING REQUIREMENTS.,...............,.................... 6-15'- o 4 26.9.'1 ROUTINE REP 0RTS........................................... 15- .! Startup Report..........................
. ................. 6-15.
v"
-Annual Reports............................................. 6-16 .l4 Annua 1' Radiological Environmental Operating Report.'....... 6-17 M 1
. Semiannual R'adioactive Effluent Release Report............ 6-18 4
# Monthly Operating Report'................................... 6-19 q Peaking Factor Limit ~ Report............................... 6-19.
-. 6. 9. 2 SPECIAL. REPORTS.......... ................................ 6-20 ,
t 6.10 RECORD RETENTION.......... ................................ 6-20
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-i TURKEY' POINT - UNITS 3 & 4 xxiii AMENDMENT N05137AND 132
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. xxiv AMENDMENT N05137AND 132 a
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1-0 AMENDMENT N05.137AND 132 j TURKEY POINT - UNITS 3 & 4 - - +v ,
t i 1.0' DEF]NITIONS l The defined terms of this section appear in capitalized type and are applicable l
. throughout these Technical Specifications. ,
ACTION 1.1 ACTION shall be that part of a Technical Specification which proscribes remedial measures required under designated conditions. MTUATIONLOGICTEST 1.2 An ACTUATION LOGIC TEST shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and i verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices. ANALOG CHANNEL OPERATIONAL TEST 1.3 AnANALOGCHANNELOPERATIONALTESTshallbetheinjectionofasimulated signal into the channel as close to the. sensor as practicable to verify . OPJRABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the setpoints are within the required range and accuracy. AXIAL FLUX DIFFERENCE j 1,4 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top-and bottom halves of a two section excore neutron detector. CHANNEL CALIBRATION- l
- - 1. 5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the required range and accuracy to known >
values of input. The CHANNEL CALIBRATION shall encompass the entire channel [. im luding the sentes and alarm, interlock and/or trip functions and may be performed by any series of sequent 41, overlapping, or total channel steps + such that the entire channel is ca * ;ted. CHANNEL CHECK
- 1. 6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior-during operation by observation. This determination shall include, where' '
possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument' channels measuring the same parameter. TURKEY. POINT - UNITS 3 & 4 1-1 AMENDMENT N05137AND 132
m g, < u . , 7,3l4ip p m. 1 sv ' . . DEFINITIONS
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,a CONTAINMENT INTEGRITY -
E 1.7' CONTAINMENT INTEGRITY shall exist when: > i a; All penetrations required to be closed during accident conditions , L , .are either:
- 1) Capable of being closed by an OPERABLE containment automatic .
p isolation valve system, or
- 1 2) Closed by manual valves, blind flan es, or deactivated automatic -
valves secured in their closed posi ions, except as provided in Table-3.6-1 of Specification 3.6.4. < j E b. The equipment hatch is closed and sealed,
- c. Each air lock is in compliance with the requirements of. Specification' 3.6.1.3,
- d. The containment leakage rates are within the limits of Specification l je 3.6.1.2, and b J e. The sealing mechanism associated with each penetration (e.g., welds, ,
[ bellows, or 0-rings) is OPERABLE.
, o CONTROLLED LEAKAGE 1.8 CONTROLLED' LEAKAGE shall be that seal water flow-supplied to the reactor :I coolant pump seals. .
CORE ALTERATIONS ;
'1.9~ CORE ALTERATIONS shall be the movement or manipulation of any component l
-within the reactor pressure vessel with the vessel head removed and fuel in 2 the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of -
movement of a component to a safe conservative position. DOSE EQUIVALENT I-131 1.10 EDOSE EQUIVALENT I-131 shall be that concentration of -131 (microcurie / gram)f ! which alone would produce the same. thyroid dose as the quantity and isotopic l mixture of I-131,.1-132, I-133, I-134, and 1-135 actually present. The thyroid-dose conversion factors used for this calculation shall be those listed in Table III of TID.-14844, " Calculation of Distance Factors for Power and Test Reactor Sites" or Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977. , s E - AVERAGE DISINTEGRATION ENERGY 1.11 E shall b9 the average (weighted in propnrtion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the ' average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample isotopes, other than iodines, with half lives greater than 30 minutes, < making'up at least 95 percent of the notal non-iodine activity in tha coolant. j TURKEY POINT - UNITS 3 & 4 1-2 AMENDMENT N05.137AND132
r , 3; ,g w , 6 DEFINITIONS f , FREQUENCY NOTATION n ' 1.12 The FREQUENCY NOTATION specified for the performance of Surveillance 1 Requirements shall correspond to the intervals defined in Table 1.1. - GAS DECAY TANK SYSTEM t "1.13 A GAS DECAY TANK SYSTEM shall be any system designed and-installed to b f reduce radioactive gaseous effluents by collecting Reactor Coolant' System off gases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the . i environment. - IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be: 1
- a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks' that are captured and conducted to a sump or collecting tank, or
- b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of Leakage Detection Systems or not to be PRESSURE BOUNDARY
- LEAKAGE, or
- c. Reactor Coolant System leakage through a steam generator to the -
Secondary Coolant System. , MEMBER (S) 0F THE PUBLIC 1.15 MEMBE.R(S) 0F THE PUBLIC shall include all: persons who are not occupa- i tionally associated with the plant. This category does not include employees ; of'the~ licensee,.its contractors, vendors or members of the Armed Forces using , property located within'the SITE BOUNDARY. Also excluded from this category.
'are. persons who enter the site to service equipment or to make Jeliveries.
This category does include persons who use portions of the site for recre- + ational, occupational, or other purposes not associated with the plant.
, OFFSITE DOSE CALCULATION MANUAL 1.16 The OFFSIft DOSE CALCULATION MANUAL (00CM) shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive '
gaseous and liquid effluents, in the calculation of gaseous and liquid effluent ronitoring Alarm / Trip Setpoints, and in the conduct of the Environ-mental Radiological Monitoring Program. h TURKEY POINT UNITS 3 & 4 1-3 AMENDMENT N05.137AND 132
+
t
+
DErlNITIONS _
+
.0PERABLE'- OPERABIllTY / :
!. l l' .1.17. A system;. subsystem, train, component or device shall be OPERABLE or l have OPERABILITY when it is capable of performing its specified fu% tion (s) .
H and when all necessary attendant. instrumentation, controls, electrical power,
< cooling or seal l water, lubrication or other auxiliary equipment that are ;
required for the system, subsystem, train, component, or device to perform'its , i function (s) are.also capable of performing their related support function (s). ]
' OPERATIONAL MODE - MODE 1 1.18 An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive I combination of-core-reactivity condition, power level, and average reactor' <
coolant temperature sperMied in Table 1.2. l PHYSICS TESTS , 1.19 PHYSICS TESTS . all be those tests performed to measure the fundamental- ' nuclear characteristics of the reactor core and related instrumentation: c (1) described in Chapter 13.5 of the FSAR, (2) authorized under the. 1 provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission, j i PRESSURE BOUNDARY LEAKAGE 1.20 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube. ; leakage) through a nonisolable fault in a Reaccor Coolant System component body, pipe wall, or vessel wall. PROCESS CONTROL PROGRAM j 1.21 The PROCESS CONTROL. PROGRAM (PCP) shall contain the current formulas, l sampling, analyses, tests, and determinations to be made to ensure that-- processing and packaging'of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplish.1 in , auch a way.as to assure compliarce with 10 CFR Parts 20, 61, and 71 ano tederal and State regulations, burial ground requirements, and other reouire- i ments governing the disposal of radioactive waste.
- . PURGE - PURGING
.1.22 . PURGE or PURGING shall be any-controlled process of discharging air or gas l
'from a confinement to maintain temperature, pressure, humidity, concentration '
or other operating condition, in such a manner that replacement air or gas is required to purify the~ confinement. ; I TURKEY POINT - UN115 3 & 4 1-4 AMENDMENT NOS.137AND 132 s 4 - - - -
DEFINITIONS , QUADRANT POWER TILT RATIO 1.23 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outputs, or the ratio of the maximum lower excore detector calibrated ,' output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used far computing the average. RATED THERMAL POWER 1.24 RATED THERMAL POWER shall be a total reactor core heat transfer rate to' the reactor coolant of 2200 MWt. REPORTABLE EVEN1 1.25 A REPORTABLE EVENT shall be any of f. hose conditions specified in Section 50.73 of 10 CFR Part 50. Ml_UTDOWNMARGIN 1.26 SHU1DOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be suberitical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn. SITEBOUNDAp 1,27 The 511E BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee. 50llDiflCAT10N 1.28 50LIDIFICA110N shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements. SOURCE CHECK 1.29 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity. STAGGERED TEST BASIS 1.30 A STAGGERED TEST BASIS shall consist of:
- a. A test schedule for n systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals, and
- b. The testing of one system, subsystem, train, or other designated component at the beginning of each subinterval.
TURKEY POINT - UNI 15 3 & 4 1-5 AMENDMENT N05.I37AND 132 l
p~.
.i DEFINITIONS :
THERMAL POWER . i 1.31 THERMAL POWER-shall be the total reactor core heat transfer rate to the reactor coolant. [ TRIP ACTUATING DEVICE OPERATIONAL TEST
- _ 1.32 A TRIP ACTUATING DEVICE OPERATIONAL TEST shall consist of operating the
-Trip Actuating Device and verifying OPERABILITY of alarm, interlock and/or '
trip functions.- The TRIP ACTUATING DEVICE OPERATIONAL TEST shall include L adjustment, as necessary, of the Trip Actuating Device such that it actuates' at the required setpoint within the required accuracy. UNIDENTIFIED LEAKAGE 1.33 UNIDENTIFIED LEAKAGE shall be all leakage which is not. IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE. UNRESTRICTED AREA 1.34 An UNRESTRICTED AREA shall be any area at ., beyond the SITE BOUNDARY-
. access to which is not controlled by the licensee fu ourpo*:s of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, '
commercial, institutional, and/or recreational purposes. VENTILATION EXHAUST TREATMENT SYSTEM 1.35 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment, Such a system is not considered to have any effect on noble gas effluents. Engineered Safety features Atmospheric Clearap Systems are not considered to be VENTILATION EXHAUST TREATMENT-SYSTEM components. VENTING , 1.36 VENTING shall be the cont rolled process of discharging air or gas from a cohfinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such t. manner that replacement air or gas is not pro-
;vided or required during VENTING. Vent, used in system names, does not imply a-VENTING process, i
TURKEY POINT - UNITS 3 & 4 1-6 AMENDMEhT N05.137AND 132 ! u 2
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- TABLE 1l1 P s ,1 -
FREQUENCY NOTATION NOTAT1014 FREQUENCY i f Si At least-once.per,12 hours. > l
,, D. At least once per 24; hours. .
- i p, .W. At least once.per;7 days. ,
- p. '
M At least.once per 31 days.- .l o , , , 3 Q, 'At least;once.per & days.. ,
- u. <
SA' At least once per 184 days.. , i 1
'R- . At least once per.18 months. <
j S/U- Prior to each reactor startup. !! a
= N. A. Not applicable. f P Completed prior-to each release. zI e
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, OPERATIONAL MODES ;y REACTIVITY '.;% RATED- AVERAGE COOLANT-h MODE ~ ' CONDITION, K THERMAL ~ POWER
- TEMPERATURE- -
eff. i., ;
- 1. POWER OPERATION: -
> 0.99 '> 5% > 350'F'
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> 0.99- 5 5% l> 350'F- 1
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- 4. HOT SHVTDOWN' < 0.99 0 -
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(' W :L @ff 6 hhd,b,MTURKEY POINTY JUNITS 3 & 4 2-0 - AMENDMENT N05137AND_132' N e@Qg ' . %~- S.*'Q rn. $ %p pi ;p. e u g4 bmd ; T g yu..e c-c <. s 1 UD
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e f" w ,
, 2.0 SAFETY LIMITS AND 'lMITING l SAFETY SYSTEM SETT'INGS 1 2.1 SAFET LIMITS' REACTOR CORE. ,
i 2.1.1 TheTcombination of THERMAL POWER, pressurizer pressure, and th'e highest- l operating' loop coolant temperature (T,yg) shall not exceed the limits shown in' y Figure 2.1-1, for 3 loop operation. APPLICABILITY: MODES I and 2. r ' L ACTION: ,
. .Whenever the' point. defined by the combination of the highest operating loop-h
~
. average temperature and THERMAL POWER has exceeded the appropriate pressurizer ;
pressure line, be in HOT STANDBY within I hour, and comply with the require- . l>< ments of Specification 6.7.1. { REACTOR COOLANT SYSTEM PRESSURE !
.2;1.2 The, Reactor Coolant System pressure shall not exceed 2735 psig. ,
APPLICABILITY: MODES 1, 2, 3, 4, and 5. , ACTION:: MODES I and 2: Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant' System pressure within its limit
, within I hour, and comply with'the requirements of Specification 6.7.1.
?
MODES 3, 4 and 5:
- Whenever the Reactor Coolant System pressure has exceeded 2735 psig, !
. reduce the Reactor Coolant System pressure to within its limit within !
5 minutes, and comply with the requirements of= Specification 6.7.1. I I' TURKEY POINT -. UNITS 3 & 4 2-1 AMENDMENT N05.137AND 132 i
. . ~ . . . . , _ _ , .
Fr 675 665 2400 PSR NI
$33 N u UNACCEPTABLE '
N 2250 PSR l
'N OPERATION g3 N .
h N i N ' ra5 M00 PSR - ' N N N T N e ,,5 N N<
\
r N 1825 PSIA g f N N 615 \ L N
~
N N N N A
=
N N 595 N A
\
585 i 575 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 1.1 1.2 POWER (FRACTION Of NOMINAL) l FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - THREE LOOPS IN OPERATION i TURKEY POINT - UNITS 3 & 4 2-2 AMENDMENT N05.137AND 132 I
.,e , - - , - - - , - - - -w .-- e-
6 'r ,
< 2, i
e ! l SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS j
- 2. 2 LIMITI$6 SAFETY SYSTEM SETTINGS, 1 1
REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS ; 4 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall. l be set consistent with the Trip Setpoint values shown in; Table 2.2-1. ! APPLICABILITY: As shown for each channel in Table 3.3-1.- ' i ACTION:_ l
- a. With aLReactor Trip System Instrumentation or Interlock Setpoint. .
.less conservative than the value shown in the Trip Setpoint column i but more conservative'than the value.shown in the A110wab~1e_Value j r
colamn of Table 2.2-1, adjust the- setpoint consistent with the Tripi ti setpoint value. l b.- With the Reactor Trip System Instrumentation.or Interlock Setpoint II less conservative than the value shown in the Allowable Values - ' column of Table 2;2-;, declare the channel inoperable and apply the ! applicable ACTION statement requirement of. Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint. adjusted- el consistent with the Trip Setpoint value. , 7 I i 4 s 4 TURKEY POINT - UNITS 3 & 4 2-3 AMENDMENT N05.137AND 132
~~
~
TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Th TRIP SETPOINT ALLOWABLE VALUE #
$ FUNCTIONAL UNIT _
5 1. Manual Rea c tor Trip N.A. N.A. c 2. Power Range, Neutron Flux
$ a. High Setpoint 1109% of RTP** 1[ ]% of RfP**
m w b. Low Setpoint $25% of RTP** $[ .]% of RTP**
- 3. Intermediate Range, 125% of RTP** $[ ]% of RTP**
Neutron Flux.
- 4. Source Range, Neutron Flux $105 cps $[ ] x 105 cps
- 5. Overtemperaturs $! See Note 1 m
A 6. Overpower AT See' Note 3
- 7. Pressurizer Prtssure-tow . >1835 psig >[ ] psig.
- 8. Pressurizer Pressure-High $2385 psig $[ ] psig
- 9. Pressurizer Water Level-High $92% of instrument span $[ ]% of instrument span
> 10. Reactor Coolant Flow-Low >90% of loop. >[ -]% of loop M Besign flow
- Besign flow
- 5 >[ .]% of narrow range instrument M 11. Steam Generator Water >15% of narrow range 5 level Low-Low Instrument-spara span 5
P-O 3E o C m
" Loop design flow = 89,500 gpm
**RTP = RATED THERMAL POWER
_ _ - _ . . = _ _ -
.-- ~ , gr =
3-
- ~ = =-~ --
n c.- mm , p.
' ~ '
. 4 5% .d
- d g :
~
'. v l .&
TABLE 2.2-1 (Continued)
~#.
is 'REAC' TOR TRIP ~ SYSTEM INSTRUMENTATION TRIP SETPOINTS x 3 FUNCTIONAL' UNIT TRIP-SETPOINT ALLOWABLE VALUE # - 5 12. Steam /Feedwater Flow . Feed Flow Feed Flow ~
- . ' Mismatch. <0.64 x 106 .lb/hr <[- -] x 106 lb/hr c Coincident With 6elow steam flow -6elow-steam flow 25 -
.d Steam Generator Water >15% of narrow- >[-' ]% of narrow range instrument w Level-Low range instrument span . span c=
.ta
- 13. Undervoltage - 4.16 kV >2496~ volts- . >[ -] volts-Busses A and B each bus .each bus
- 14. Underfrequency - Trip of Reactor >56.1 Hz'
>[
] Hz Coolant Pump Breaker (s) Open m
J, 15. Turbine Trip
- a. Auto Stop Oil-Pressure >45 psig >[ ] psig -
- b. Turbine Stop Valve Fully Closed *** Fully Closed ***
~
Closure
- 16. Safety Injection Input- N. A. N.A.
> from ESF M
5 17. Reactor Trip System M Interlocks 5 2 a. Intermediate Range '>1.x 10 10 amp. ->[' -] amp L 8 Neutron Flux, P-6 -- t$ E o a .
^^^ Limit. switch is set when Turbine Stop Valves are fully closed.
. . . . _ _. _ . _ . . _ _ . . __; i,. _.;_ _ _ _ _ =_ __12 . ._ _. _ . _ _ . . __ z _1 _ _ _ = .
'^ - -
~
., 2.+ .
. ~-
.. ,a
~
TABLE 2.2-1 (Continued) _ . . ..
'E- REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.
m FUNCTIONAL UNIT ' TRIP SETPOINT- -ALLOWABLE VALUE #
'T b. Low Power Reactor Trips' -
Block, P-7
' ~~
7 E 1) P-10 input <10% of RTP**'
-o
'$[ ]% of RTP** .
- 2) Turbine first Stage $10% Turbine Power .$[ ]% Turbine Power- 1 Pressure
- c. Power Range Neutron $45% of RTP** '<[ .]% cf RTP**'
Flux, P-8.
- d. Power Range Neutron >10% of RTP** >[ ]% of RTP**
Flux, P-10 7
- 18. Reactor Coolant Pump .N.A. N.A.
Breaker Position Trip ..
- 19. Reactor Trip Breakers N.A. N.A.
- 20. Automatic Trip and Interlock N.A. N.A.
Logic E 9 E . 2 5 m
^
E! E o --
,3 -.
n3 . _.
,i
**RTP = RATED THERMAL POWER -
.i
_ ,y
. - - . . _ . --.l.'.-
~
- n ,
;. n- .- , - . .,~ ;- -- r-m L. ~ ", -
,._ ~ ~ -.
TABLE 2.2-1 (Continued)~ - -
'U. TABLE NOTATIONS =
EE X NOTE li DVERTEMPERATURE AT b 1 AT ( ) < AT {K - K . U + T25) [T.( 1 ) - T'] + K-(Pj-.P') - f (aI)}. 5 1+t5 ~ o. I 1 2 (1.+ t 35) 1+1 45 3 e _ C
?5
--e Where: AT = ' Measured aT by RTD Instrumentation (includes RTD response time);.
1 w 1 ,3
= Lag compensator on measured AT;-
e Ti = Time constants utilized in the lag compensator for AT, = t, + RID. response time = 2.5s; AT g = Indicated tJ at RATED THERMAL POWER (includes RID response time);
'K i = 1.095;
'if K2 = 0.0107/*F; 1+1 2 =
. The function generated by the lead-lag compensator for T,yg.
dynamic compensation;-
=
r2 Is Time constants utilized in the lead-lag compensator,for Tavg12.= 25s,.- T3 = 3 s; g .T = Average temperature, *F; Z l = k 1+I54 Lag compensator on measuM T,yg; 5 5 14 = Time constant utilized in the measured T avg lag compensator, S .- 14 + RID response time = 2.5s;
> T' <
574.2 F (Nominal Tavg at RATED THERMAL POWER); z _ o
- m K3 = 0.000453/psig; w
N P = "ressurizer pressure, psig; -
~
y y- v. m,- .Agrbg---Try,. 3' eye-.. irdyp.g H4 +gy-n *P9'g y P *-M**g *3=p** w + =t4*s' = ' WM aM/'Q- v e *?r'FWe'c* N--t*-'m-'I'T y'FY th** vN**e 'T WMPWs'**$'F*4*TMt- M *'**W.-.W*1'- '"**'W'n'* *'M#"*N 4 t'-TT
- M
n _. , . , _ y z* . _
. +
- . sv-
~
~
TABLE 2.2-1 (Continued) TABLE NOTATIONS (Continued) [ NOTE 1: (Continued).
@ P' = 2235 psig (Nominal RCS operating pressure); '
-5 '= Laplace transform operator,'s 2;
[
~
and f-(aI) is a function of the indicated difference'between top and bottom detectors of the-
" power-range neutron ion chambers; with gains to be selected based on measured instrument-response during plant startup tests such that:
(1)' for q't - gb between - 14% and + 10%, f'(AI) = 0, where g and g a are b percent RATED TrcRMAL" POWER in the top and bottom halves of the core respectively, and q t + 3 s total-THERMAL POWER in percent of RATED THERMAL POWER; (2) For each percent that the magnitude of _q t- % exceeds - 14%, the AT. Trip Setpoint shall . d> be automatically reduced by 2.0% of its value at RATED THERMAL POWER; and (3) For each percent that the magnitude of q - gyexceeds+.10%.~theATTripSetpointshall be automatically reduled by 3.5% of its value at RATED THERMAL POWER. NOTE 2: (This note number is.not used.) g ,- M
*i . .
2 S.
~
/
.- g
~
ro w h-L -_vkre
- fr'-th-@ 9- _- S **-CT' 4Au W MW'@F 'S' & E-'P*7 -"hO7 W- '+='**W 'IP *t '1*F--h *- e '?'#W '"**@TilfW'Ne'rM* * -N'*---- -.sN* =.3e' wT emt*En air' =# '* * 'Elke' h4=-t*'h- ' ' = = = * - -= ^--- - ----- --
w; -
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^
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- ~
TABLE 2.2-1 (Continued)-
- a ~
J=. =W.m;?" t
} "
TABLE NOTATIONS"(Continued) # 2[ ~
, : r .J: ~ 5 3 NOTE 3; OVERPOWER aT
~ ~ '
E -
] AT ( 1 ( T35' ) (- 1 )T-K [T I I
) < AT {K4 -K I - T"I- f (aI)} -
c- (1 + .t iS) ~ o 5 (1 + 13 5) (1 + 1 45) S
.(1 + T4 5) 5 3
w Where: AT = As defined in Note 1, 1 a y, 3
= As defined in Note 1, 1 = As defined in Note 1, AT, = As defined in Note 1 -
ru 4 K4 = 1.09, Ks = 0.02/*F for increasing average temperature and 0 for decreasing. average temperature, I5 = _1 fs3 The function generated by the rate-lag' compensator for T,,9 Wic compensation,
> =
13 Time constants utilized in the rate-lag compensator for Tavg, r3 = 10 s, x I
= As defined in Note 1, "h
1+T 45 E w z 14 = As defined in Note 1, 8 3 I h.
- m e
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i 2.1 SAFETY LIMITS ; BASES-2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in.the release of. fission , products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the j heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature. ! Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure j from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer t coefficient, DNB is not a directly measurable parameter during operation and , therefore THERMAL POWER and reactor cool 6nt temperature and pressure have been ! related to DNB. This relationship has been developed to predict the DNB flux , and the' location of DNB'for axially uniform and nonuniform heat flux distribu-tions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that-would cause DNB at a particular core location to the local heat .' flux and is indicative of the margin to DNB. , The DNB design basis is as follows: there must be at least a 95-percent : probability with 95 percent confidence that the minimum DNBR of the limiting rod during Condition I and 11 events is greater than or equal to'the DNBR
- linit of the DNB correlation being used. The correlation DNBR. limit is ,
established based on the entire applicable experimental data set such that t there is a 95 percent' probability with 95 percent confidence that DNB will - not occur when the minimum DNBR is at the DNBR' limit. The cu,ves of Figure 2.1-1 show the loci of points of THERMAL POWER, ; Reactor Conlant System pressure and average temperature for which the minimum DNBR is no less than the design DNBR value, or the average enthalpy at the vessel eW t is equal to the enthalpy of saturated liquid. These-curves are based on an enthalpy hot channel factor, F of 1.62 and , andar(ferencecosinewithapegkof,1.55foraxialpowershape.H,Anallowance ' is included for an increase in F3g at reduced power. based on the expression: F H 5 1.62 (1+ 0.3 (1-P)] Where P is the fraction of RATED THERMAL POWER. These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion limit assuming the axial power imbalance is within the limits of the f (AI) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-temperature AT trips will reduce the setpoints to provide protection consistent with core Safety Limits. t TURKEY POINT - UNITS 3 & 4 B 2-1 AMENDMENT N05.137AND 132
n n [,, SAFETY LIMITS F BASES' ['
; 2.' 1.1 REACTOR CORE-(Codinued)
Fuel rod bowing reduces .the values of DNB ratio (DNBR). The penalties'are
-calculated pursuant to." fuel Rod Bow Evaluation," WCAP-8691-P-A Revision 1- , (Proprietary) and WCAP-8692 Revision 1 (Non-Proprietary). . The restrictions of the Core Thermal Hydraulic Safety Limits assure that an amount of DNBR margin greater.than or equal to the above penalties is retained to offset the rod bow DNBR penalty.
L' I 2.1.2' REACTOR ~ COOLANT SYSTEM PRESSURE The restriction'of this Safety Limit protects the. integrity of the Reactor-Coolant System (RCS) from-overpressurization and thereby prevents the release g of radionuclides contained in-the reactor coolant from reaching the containment ' atmosphere.- The-reactor vessel and pressurizer are designed to Section III of the ASME 7' Code for Nuclear. Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The RCS piping, valves and fittings are designed to ANSI B31.'1 which permits a maximum transient pressure of 120% of' design pressure of 2485 psig. The Safety Limit of 2735 psig is therefore more conservative than the ANSI B31.1 design criteria and consistent with associated ASME Code requirements. The entire RCS is'hydrotested at 125% (3107 psig) of design pressure, to
. demonstrate integrity prior to initial operation.
s JTURKEY POINT. - UNITS 3 & 4 B 2-2 AMENDMENT NOS.137AND 132 o m
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^ 2;2 LIMITING SAFETY SYSTEM-SETTINGS E ;
i BASES
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li . I 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SCTPOINTS n , f The~ Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the core and Reactor Coolant
-System are prevented from exceeding their safety. limits'during normal operation and design basis anticipated operational-occurrences and to assist the-Engi-p neered Safety Features Actuation System in mitigating the consequences of l p accidents.
N To accomn.odate the instrument drif t that may occur between. operational 1 tests and the accuracy to which setpoints can be measured and calibrated,
; Allowable Values for the Reactor Trip Setpoints have been specified in Table'2.2-1. . Operation with a trip set.less conservative than its Trip
'Setpoint but within its specified Allowable Value is acceptable. If no value is listed in the Allowable column, ^he setpoint value is the limiting setting.
The methodology to derive the Trip Setpoints includes an allowance for l
. instrument uncertainties. Inherent to the determination of the Trip Setpoints ;
are the. magnitudes of.these channel uncertainties. Sensors and other instru-
,' mentation utilized in these channels are expected to be capable of operating e within the allowances of these uncertainty magnitudes. !
1 The various Reactor trip circuits automatical_1y open the Reactor trip r
, breakers whenever a condition monitored-by the Reactor Trip System reaches a -
preset or calculated level. In addition to redundant channels and trains, the. desigr. approach provides a Reactor Trip System which monitors numerous system , variables, therefore providing Trip System functional diversity. The functional capability at the specified trip setting is required for those anticipatory or ; diverse Reactor trips for which no direct credit was assumed in the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a Turbine-trip signal whenever Reactor trip is initiated. This prevents the reactivity insertion that would otherwid result i from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation. System. l Manual-Reactor Trip , The Reactor Trip System includes manual Reactor trip capability. TURKEY P01'NT - UNITS 3 & 4 B 2-3 AMENDMENT N05.137AND 132 ,
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lf y. f xy-LIMITING-SAFETY SYSTEM SETTINGS T- BASES I Power Range, Neutron Flux o
-In each of the Power Range Neutron Flux channels there are two independent s bistables, each with its >>wn trip setting used for a High and Low Range trip setting. The low Setpoint trip provides protection during suberitical and low Z
p power operations to mitigate the consequences of a power excursion beginning j 1
- from low power, and the High Setpoint trip provides protection during power 1 l operations for all power levels to mitigate the consequences of a reactivity excursion which may be too rapid for the-temperature and pressure protective
{, trips. N. . The Low Setpoint trip may be manually blocked above P-10 (a power level lF of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.- Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncor-trolled rod cluster control assembly bank withdrawal from a subtritical
- condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will
-initiate a Reactor trip at about 105 counts per second unless manually blocked 4
, when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit
'is taken for operation of the trips associated with either the' Intermediate or Source-Range Channels in the accident analyses; however, their functional.
'p capability at the specified trip settings is required t.y this specification to enhance the overall reliability of the Reactor Protection System.
[ .0vertemperature .iT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial psver p ' distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (ebout 4 seconds), . and pressure is within the range between the Pressurizer High and Low Pressure trips, The setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water , and includes dynamic compensation for piping delays from the core to the loop. I - ' s temperature detectors, (2) pressurizer pressure, and (3) axial power distribu-s , tion. With normal axial power distribution,.this Reactor trip limit-is always i' below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are 6 greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1. Table 2.2-1, Note 1, provides the equation defining overtemperature AT. Inherent in the equation are time delays on measured AT and T gg attributable to RTDs and lag compensators. A time delay of 2.5 seconds is introduced through a combination of RTD response : time and adjustment of time constants t i and 14 _. TURKEY POINT - UNITS 3 & 4 B 2-4 AMENDMENT N05.137 AND 132 i
7 s [ ;c LIMITING _ SAFETY SYSTEM SETTINGS , BASES Overpower AT i The Over exceeding 118% power of the AT power design trip prevents density.power density anywhere This.provides in of assurance the core from fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) e
' under all possible overpower conditions, limits the required range for Over- -
temperature AT trip, and provides a backup to the High Neutron Flux trip. The ! setpoint is automatically varied with: (1) coolant temperature to correct.for L temperature induced changes in density and heat capacity of water, (2) rate of ' change of temperature for dynamic compensation for piping delays from the core ; to the-loop temperature detectors, and (3) axial power distribution, to ensure f that the allowable heat generation rate (kW/ft) is not exceeded. Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in-the event of a loss of reactor coolant pressure. On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine first stage pressure at 'approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7. TheHighSetpointtrip'functionsinconjunctionwiththepressurizer safety valves to protect the Reactor Coolant System against system overpressure.
- Pressurizer Water Level The Pressurizer Water Level-High trip is provided to prevent water relief through'the pressurizer safety valves. On decreasing power the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine first stage pressure at approximately 10% of full power equivalent); and on increasing power, auto- ,
matica11y reinstated by P-7.
' Reactor Coolant Flow The Reactor Coolant Flow-Low trip provides core protection to prevent DNB
- by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.
On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine first stage pressure at approximately 10% TURKEY POINT - UNITS 3 & 4 B 2-5 AMENDMENT N05137AND 132 I er
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LIMITING SAFETY SYSTEM SETTINGS i BASES-Reactor Coolant Flow (Continued) - o !. of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of loop design flow. Above P-8 (a power-level of approximately 45% of RATED THERMAL POWER) an automatic Reactor trip- , L . will occur if the flow in any single loop. drops below 90% of nominal full loop ' flow. Conversely, on decreasing power between P-8 and the P-7 an automatic l- Reactor. trip will occur on low reactor coolant flow'in more than one loop and - [ below P-7 the trip function is automatically blocked. j: Steam Generator Water level , The Steam Generator Water Level Low-Low trip protects the reactor from
~1oss of heat sink in the event of a sustained steam /feedwater flow mismatch .
resulting f rom loss of normal feedwater. The specified setpoint provides i allowances for starting delays of the Auxiliary Feedwater System. Steam /Feedsater Flow Mismatch and Low Steam Generator Water level i The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generatcr Water Level-Low trip is not used in the transient and accident analyses but is 3 included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Trip System. This trip is redundant to the Steam Generator Water Level Low-Low ' trip. The Steam /Feedwater Flow Mismatch portion of this trip is activated > when the steam flow exceeds the feedwater flow by greater than or equal to 0.64 x 100 lbs/ hour. The Steam Generator Water Level-Low portion of the trip 1 is activated when the watar level drons below 15%, as indicated by the narrow range instrument. These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a Reactor
. trip before the steam ge.'erators are dry. Therefore, the required capacity and starting time requirements of the auxiliarf feedwater pumps are reduced and the resulting thermal transient on the Reu. tor Coolant System and steam generators is minimized.
Undervoltage - 4.16 kV Bus A and B Trips The 4.16 kV Bus A and B Undervoltage trips provide core protection against DNB as a result of complete loss of forced coolant flow. The specified setpoint assures a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time delays are incorporated in the Undervoltage trips to prevent spurious Reactor trips f rom momentary electrical power transients. The delay is set so that the time required for a signal to reach the Reactor trip breakers following the trip of at least one undervoltage relay in both of the associated Units 4.16 kV busses shall not exceed 1.3 seconds. On decreasing TURKEY POINT - UNITS 3 & 4 B 2-6 AMENDMENT N05.137AND 132 i
C z ' LIMITING SAFETY SYSTEM SETTINGS m l 1 BASES-L j Undervoltage and :- 4.16'kV Bus A and B ' rips (Continued) , power the Undervoltage Bus trips are automatically blocked by P-7 (a power level of approximately 10%-of RATED THERMAL POWER.with a turbine first stage ; pressure at approximately 10% of full power equivalent); and on increasing i power, reinstated automatically by P-7. j Turbine Trip I A Turbine trip initiates a Reactor trip. On decreasing power, the Reactor Trip from the Turbine trip is automatically blocked by P-7 (a power level of approximately 10%'of RATED THERMAL POWER with a turbine first stage pressure at approximately 10% of full power equivalent); and on increasing power, reinstated automatically by P-7. ; Safety Injection Input from ESF I If a. Reactor trip has not already been generated by the Reactor Trip-System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection. The ESF instrumentation channels which initiate a Safety Injection signal are
.shown in Table 3.3-3.
Reactor Coolant Pump Breaker Position Trip l . The Reactor Coolant Pump Breaker Position ' rips are anticipatory trips which provide reactor core protection-against DNB. The open/close position trips assure a reactor trip signal is generated before the low flow trip ! setpoint is reached. No credit was taken in the accident analyses-for operation
- of these trips. Their functional capability at the open/close position settings ,
is required to enhance the overall reliability of the Reactor Protection System. Above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a
'o turbine first stage pressure at approximately 10% of full power equivalent) an automatic reactor trip will occur if more than one reactor coolant pump breaker
'is opened. Above P-8 (a power level of approximately 45% of RATED THERMAL POWER) an automatic reactor trip will occur if one reactor coolant pump breaker ,
is-opened, On decreasing power between P-8 and P-7, an automatic reactor trip ; will occur if more than one reactor coolant pump breaker is opened and below , P-7 the trip function is automatically blocked. Underfrequency sensors are also installed on the 4.16 kV bussas to detect underfrequency and initiate breaker trip on underfrequency. The underfrequency trip setpoints preserve the coast down energy of the reactor coolant pumps, in case of a grid frequency decrease so DNB does not occur, i TURKEY POINT - UNITS 3 & 4 8 2-7 AMENDMENT N05.137 AND 132 ,
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( -- i Reactor Trip System Interlocks [ The Reactor Trip. System interlocks perform the following functions: P-6 On increasing power, P-6 allows the manual block of-the' Source Range g trip (i'.e. , prevents premature block of Source Range trip) and
" deenergizes the high voltage to the detectors. Ondecreasingpowert
' Source Range level trips are automatically reactivated and high voltage estored, b
N P-7 On' increasing power, P-7 automatically enables. Reactor trips on low. l' flow in more than one reactor coolant loop, more than one reactor. coolant pump breaker open, reactor coolant pump bus undervoltage and underfrequency, Turbine trip, pressurizer low pressure and pressurizer high level.- On decreasing power, the above listed trips are auto-matica11y blocked.
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P-8 On increasing power, P-8 automatically enables. Reactor trips on low-
- l. flow in one or more reactor coolant loops, and one or more reactor coolant pump breakers open. On decreasing power, the P-8 interlock'
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automati n lly blocks the trip on low flow in one coolant loop-or one coolant pump breaker open. . If J a P-10 On inc.* easing power, P-10 allows the manual block of the Intermediate L Range trip and the Low Setpoint Power Range < trip; and automatically. blocks the-Source Range trip and deenergizes the Source Range high voltage power. On decreasing power, the Intermediate Range trip and the Low Setpoint Power Range trip are automatically reactivated. P-10 also provides input to P-7. The trip setpoint on increasing power shall be > 10% and the reset point shall be less than or equal-to 10%. L. i y L zTURKEY POINT --UNITS ? & 4 8 2-8 AMENDMENT N05 137AND 132 :, 1 I
l e 1 4. I '. 7 SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS.FOR OPERATION AND SURVEILLANCE REQUIREMENTS I i i
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l TURKEY POINT - UNITS 3,& 4 3/4 0-0 AMENDMENT N05.137 AND 132
( l y l 3/4 ' LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS-
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3/4.0 APPLICABILITY LIMITING CONDITIONS FOR OPERATION ' T:: ~.~:~. :.. ., 3.0.1 Compliance with the Limiting Conditions for 0parMi- o etained in the succeeding specifications-is required during the OFDW EAL MODES or , other conditions specified therein; except that upon failure to mset the Limiting Conditions for Operation, the associated ACTION reouire'.ients shall j be met. ' 3.0.2 Noncompliance with a specification shall exist when the requirements of
' the Limiting Condition for Operation and associated ACTION requirements are '
not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time interva h , completion of the ACTION requirements is not required. 3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour action shall be initiated to place the unit, as applicable, in:
- a. -At least HOT STANDBY within the next 6 hours,
- b. At least HOT SHUTDOWN within the following 6 hours, and
- c. At laast COLD SHUTDOWN within the subsequent 24 hours.
Where corrective measures are completed that permit operation under the ACTION requirements, the action may be taken in accordance with the specified time ; limits as measured from the time of failure to meet the Limiting Condition for ! Operation. Exceptions to these requirements are stated in the individual specifications. This specification is not applicable in MODES 5 or 6. 3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made when the conditions for the Limiting Conditions for Operation are not , met and the associated ACTION requires a shutdown if they are not met within a specified time interval. Entry into an OPERATIONAL MODE or specified condition may be made in accordance with ACTION recuirements when conformance to them permits continued operation of the facility for an unlimited period of time. This provision shall not prevent passagn through or to OPERATIONAL MODES as required to comply with ACTION requirements. Exceptions to these requirements are stated in the individual specifications. , TURKEY POINT - UNITS 3 & 4 3/4 0-1 AMENDMENT N05.137AND 132
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'? LIMITING CONDIT10NS'FOR:0PERATION (Continued)- :;
y 4 s ly , s ; Il. M. 3.0.5' Limiting Conditions for Operation including lthe associated ACTION-1
-requirements shallfapply'to;each unit' individually unlessiotherwise indicated.
as.follows: 3 m- . . . .
-a-- Whenever the Limiting Conditions'fo'r Operation refers:to systems or- - .;
, , ' components'which are shared by both_ units, the ACTION requirements zc u will apply to both units simultaneously.' '}
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,' b .- Whenever .the Limiting Conditions for Operation ~ applies; to only.one- :!
unit, this will be identified .in the- APPLICABILITY section- of- the;
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' speci f,ication; ' and !
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Whenever certain portions'of a specification contain operating
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i parameters, Setpoints, etc., which:are different.for each unit,'this ! will be identi_fied in parentheses, footnotes or body of the J} requirement. J
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. TURKEY POINT - UNITS 3 & 4 3/4 0-2 AMENDMENT N05.137AND 132
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-APPLICABILITY" SURVEILLANCE REQUIREMENTS e
4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for
, , Operation unless otherwise stated in an individual Surveillance Requirement.
' 4. 0. 2 Each Surveillance Requirement shall be performed within the specified
, time interval with a maximum allowable extension not to exceed 25% of the surveillance interval.
4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliansa with the OPERABILITY requirements for a Limiting Condition for Operation. The time limits of the ACTION requirements are applicable at-the time it is identified that a Surveillance Re.quirement has not been performed. The ACTION requirements may be delayed for up to 24 hours to permit th'e completion of the surveillance when the allowable outage time limits.of the ACTION requirements are less than 24 hours. Surveillance Requirements do not have to be performed on inoperable equipment. 4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the_ Surveillance Requirement (s) associated with a Limiting e Condition for Operation has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or te OPERATIONAL MODES as required to comply with ACTION requirements.
, 4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, and 3 components shall be applicable as follows:
- a. Inservice inspection of ASME Code Class 1, 2, and 3 components and-inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of.the ASME Boiler and Pressure Vessel Code and applicable Addenda as_ required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
i TURKEY POINT - UNITS 3 & 4 3/4 0-3 AMENDMENT N05.137 AND 132
y APPLICABILITY . SURVEILLANCE REQUIREMENTS (CONTINUED)-
- h. Surveillance intervals specified in Section XI of the ASME Boiler an_d Pressure Vessel Code and applicable Addenda shall be applicable as .follows in'these Technical Specifications:
ASME Boiler and Pressure Vessel Required frequencies for-Code and applicable Addenda performing inservice terminology for inservice inspection and testing inspection and testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least.once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days. Yearly or annually At least once per 366 days
- c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities,
- d. Performance of.the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements,
- e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede-the requirements of any Technical Specification.
4.0.6 Surveillance Requirements shall apply to each unit individually unless otherwise indicated as stated in Specification 3.0.5 for individual specifications or whenever certain portions of .a specification contain surveillance parameters different for each unit, which will be identified in parentheses, footnotes or body of the requirement. 1 132 TURKEY POINT - UNITS 3 & 4 3/4 0-4 AMENDMENT N05.I37AND
p, - - - p , 6 h " '3/4.1 REACTIVITY CONTROL SYSTEMS' 3/4.1.1 BORAT10N CONTROL- I O ' SHUTDOWN MARGIN. T,y GREATER THAN 200"F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal t'o'the applicable 4
- value shown in Figure 3.1-1. ,
i 1
. APPLICABILITY- PODES 1, 2*, 3, and 4.
ACTION: . ." ~With the SHUTDOWN r ..su . . -ss than the applicable value shown in Figure 3.1-1, , immediately initiate and continue boration at greater than or equal to 4 gpm of a solution containing greater than or equal to 20,000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 3
~
4.1.1.1.1' The SHUTDOWN MARGIN shall be determined to.be greater than or equal: to the-applicable value shown in Figure 3.1-1: -
- a. Within I hour af ter detection of an inoperable control rod (s) and at least once per-12 hours thereafter while the rod (s) is inoperable.
If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased , allowance for the withdrawn worth of the immovable or untrippable-control rod (s);
- b. When in MODE 1 or MODE 2 with Keff greater than or equal to 1 at least once per 12 hours by verifying that control. bank withdrawal is' within the limits of Specification 3.1.3.6;
- c. less than l', within 4 hours prior to achieving WheninMODE2withK)7herifying.thatthepredictedcriticalcontrol reactor criticality _b rod position is within the limits of Specification 3.1.3.6;
- d. Prior to initial operation above 5% RATED THERMAL POWER after each
' fuel loading, by consideration of the factors of Specification 4.1.1~.1.le.-below, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and
^5ee Special Test Exceptions Specification 3.10.1.
TURKEY 00lNT - UNITS 3 & 4 3/4 1-1 AMENDMENT N05137 AND 132
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?:6 s ei WhenLin-MODEf3 ori4~,fatEleast'. once per 24 hours l by consideration of-f :l
. the.following1 factors:.
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- 1) Reactor-Coolant System boron concentration,3 '
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- 3) Reactor Coolant S stem average temperature, '
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- 5) Xenon.concentrat-ion, and; i
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- 6) Samarium concentration.
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'f.4.liL1.2 ? When-in MODE l'or 2, the overallLcore recctivity balance; shall be-
' compared-to predicted values to: demonstrate agreement;within 1 10 4k/k-at least;
#~ once per' 31TEffecti'e v Full Power Days (EFPD). This comparison shallcconsideri' '
.atEleastithoseLfactors stated in Specification 4.1.1.1.1e,Jabove? The,
- predicted reactivity val.ues shall be adjusted'(normalized)..to' correspond'to: -
the{ actual core conditions prior to exceeding a' fuel burnup'of 60 EFPD1after? :]j.
, eachifuelfloading,-
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; TURKEY POINT - UNITS 3 & 4 3/4 1-2 AMENDMENT N05 137AND 132. !
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1.0 A (750,1,0) l W 0.5 0.0 0 $00 1000 1500 2000 RCS BORON CONCENTRATION (PPM) , Figure 3.1-1 Required Shutdown Margin vs Reactor Coolant Boron Concentration TURKEY POINT - UNITS 3 & 4 3/4 1-3 AMENDMENT NOS.i?'AND 132
y n. ~ o Qc y_x ,
~
g 'i - - REACTIVITY-CONTROL SYSTEMSJ , SHUTDOWN MARGIN -LT,yg LESSLTHAN~0R EQUAL- TO 200 F-3 ~ T: LIMITING: CONDITION FOR OPERATION
~
j
-3.1l1.2 The SHUTOOWN MARGIN shall be greater than or equal '.toi1% Ak/k.
l r APPLICABILITY: ~ MODE 5. 1; j a ACTION: [ 37'
' With the . SHUTDOWN MARGIN less-than 1% Ak/k, immediately-initiate and continue boration..at greater than or equal to 4'gpm of a solution containing greater '
- u. than or equal to,20,000 ppm' boron ~or equivalent until the required SHUTDOWN'
, : MARGIN is~ restored.' :]
m
-t
, SURVEILLANCE REQUIREMENTS .i f
, y , I L4 . 4.1.1.2' The:. SHUTDOWN MARGIN shall be determined;to be. greater than or equal l' to Ll%. Ak/k:-- .;
- a. 'Within I hour after detectior, of an inoperable control rod (s) andrat I least once per 12 hours thereafter while the rod (s) is inoperable.
~
If the inoperable; control rod is immovable or untrippable, the j SHUTDOWN MARGIN shall: be verified acceptable _ with an increased ' : allowance for the withdrawn worth of the immovable or untr.ippable. ! control' rod (s); and: t
.b. At-leas _t'once per 24-hours by' consideration.of the following factors:
1 1)- Reactor. Coolant System boron concentration, t
- 2) LControl rod _ position,
- 3) Reactor Coolan_t System average temperature, b 4) Fuel burnup based on gross thermal energy generation ~,
- 5) Xenon concentration,-and
- 6) Samarium concentration, m i 3
f TURKEY POINT - UNITS 3 & 4 3/4 1-4 AMENDMENT N05.137AND 132 r
p - 9 7 L, 7, REACTIVITY CONTROL SYSTEMS-
. MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1. 3 The moderato, temperature coefficient (MTC) shall be:
Less positive than or equal to 5.0 x 10 5 Ak/k/ F for all rods withdrawn, beginning of the cycle life (BOL), hot zero THERMAL POWER (HZP) conditions; and ,
- b. Less positive than or equal to 5.0 x 10 5 Ak/k/ F from HZP to 70%
RATED THERMAL POWER condition; and
- c. Less positive than or equal to 5.0 x 10 5 Ak/k/ F from 70% RATED THERMAL POWER. decreasing linearly to less positive than or equal to -l 0 Ak/k/ F at-100%' RATED THERMAL POWER conditions; and !
i
- d. Less negative than -3.5 x-10-4 Ak/k/ F for the all rods withdrawn, ]
end of cycle life (E0L), RATED THERMAL POWER condition. < i APPLICABILITY:-. Specification 3.1.1.3a, b and c. - MODES 1 and 2* only**. l
. Specification 3.1.1.3d. - MODES 1, 2, and 3 only**. ]
ACTION:
- a. With the MTC more positive.than the limit of Specification 3.1.1.3a, b or c above, operation in MODES 1 and 2 may proceed provided:
1
- 1. Control rod withdrawal limits are established end maintained 1 sufficient to restore the MTC to less.poeth ve cr equal to j limits described in 3.1.1.3a, b and c abave within 24 hours or l be.in HOT STANDBY within the next 6 hours. These withdrawal- -l limits shall be in addition to the insertion. limits of-Specification 3.1.3.6; j
- 2. The control rods are maintained within the withdrawal limits i established above until a subsequent calculation verifies that l the MTC has been re4tored to within its limit for the all rods withdrawn condition; and {
q
- 3. A Special Report is prepared and submitted ,ot the Commission, j pursuant to Specification 6.9.2, within 10 days, describing the value of the measured MTC, the interim control rod withdrawal l limits, and the predicted average core burnup necessary for i restoring the positive MTC to within its limit for the all rods- s withdrawn condition. 4
*With K,ff greater than or equal to 1.
**See Specia'l Test Exceptions Specification 3.10.3. ]
ll 1 TURKEY POINT - UNITS 3 & 4 3/41-5 AMENDMENT N05.137 AND 132
$?n n -z e . ,
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= REACTIVITY CONTROL SYSTEMS'
' LIMITING-CONDITION FOR OPERATION:
k -
.ACTIONi (Continued)L
;c b.- With the MTC more negative than.the limit;of Specification 3.1.1.3di..
- above, be in HOT SHUTDOWN within-12 hourst
-SURVEILLANCE REQUIREMENTS
__e
. I 4.1.1.3~ The MTC shall.be determined to be within-its limits during each fueli ;
cycle:as follows:-
- a. The MTC.shall.be measured and~ compared-to the BOL limit'of Specifi-K' Jcation 3.1.1.'3ac, above, prior to initial operation above 5% of.
RATED. THERMAL POWER, after each fuel loading; and - b., The MTC sh'all be' measured at any -THERMAL POWER and. compared to- "t
-3,0 x.10 4 ok/k/ F (all rods withdrawn, RATED THERMAL POWER; condition) within 7 EFPD af ter. reaching an equilibrium boron'concen-tration of. 300 ppm. In the event this comparison indicates the MTC is more negative than.-3.0 x 10 4 ok/k/ F, the MTC.shall' be i.
remeasured, and compared to the.E0L MTC limit of Specification
-3.1~.1.3d. , at leastLonce per 14 EFPD during the remainder of the
-fuel cycle. -
L , l
- c. Performidesign calculation to verify -conformance;to- Specifications
, 3.1.1.3b and c. _ . l a
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-TURKEY POINT - UNITS 3 & 4 3/4 1-6 AMENDMENT N05.137 AND 132 o
x REACTIVITY CONTROL-SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY
-LIMITING CONDITION FOR OPERATION 13.1.1.4 The Reactor Coolant System lowest operating loop temperature' (Tavg) shall be greater than or equal to 541 F.
APPLICABILITY: MODES I and 2* **. ACTIONi With a Reactor Coolant System operating loop temperature (Tavg) less than
-541 F, restore T to within its limit within 15 minutes or be in HOT 3yg STANDBY within the next 15 minutes.
SURVEILLANCE REQUIREMENTS 4.1.1.4 The React'or Coolant System temperature (Tavg) shall be determined to be greater than~or equal to 541 F:
.a. -Within 15 minutes prior to achieving reactor criticality, and
- b. At.least once:per 30 minutes when the reactor is critical and the-Reactor Coolant SystemasT 'g is less than 547 F-with the T avg T7ref Deviation Alarm not rac.
*With KeH greater than or equal to 1.
, **See Special Test Exceptions Specification 3.10.3.
. TURKEY POINT - UNITS 3 & 4 3/4 1-7 AMENDMENT N05 137AND 132
p, , , p m g (3 i-s>g h REACTIVITY CONTROL, SYSTEMS; \ h 3/4.1.2 BORAT10N SYSTEMS-p & V > FLOW PATH ' SHUTDOWN- - , LLIMITING CONDITION FOR-0PERATION-y" '3.1.2.1' As a minimum,'one'of the,following boren' injection flow paths shall-
~be OPERABLE,and capable of being powered from an OPERABLE emergency power source:
- a. 'A flow path from the bori.c acid storage' tanks via:a boric acid '.
4 transfer pump.and a charging pump to the Reactor Ccolant System-if. ' l the boric acid storage ' tank in Specification 3.1.2.4a. -is OPERABLE, or
, b. The' flow path from the refueling water storage tank viaLa charging.
h pump to~-the Reactor-Coolant System if the refueling water storage tank- in Specification 3.1.2.4b. is ~ 0PERABLE.' t APPLICABILITYi MODES 5 and 6. -j E ' ACT10N: ]
.With;none of the above flow paths OPERABLE or capable of being powered from ani
'] .
j 4 OPERABLE emergency power source, suspend'all operations involving CORE? l
' ALTERATIONS or' positive reactivity changes, ,
. SURVEILLANCE REQUIREMENTS- i L- 4;1.2.1 At,least one of the above required flow paths shall be demonstrated j OPERABLE: :-
. . a b- a. At least once per 7 days by verifying that the temperature:of the~ -)
heat traced portion of the flow path is' greater than or equal to j 145 F when a flow path from'the boric acid' tanks is used, and 1
.i n b. At least once per 31 days by verifying that each valve (manual, . j power-operated, or automatic) in the, flow path that is not locked, 1
sealed, or otherwise' secured in position, is in its' correct
~
. position. .
l 1 TURKEY POINT - UNITS 3 & 4 3/4 1-8 AMENDMENT N05.137AND 132 q
em , , y ( , , p' :m , REACTIVITY CONTROL' SYSTEMS-
- FLOW PATHS - OPERATING LIMITING CONDITION-FOR OPERATION 3.1.2.2 The following boron injection flow paths shall be OPERABLE:
p a . The source path'from a boric acid storage tank via a boric acid transfer pump to the charging pump suction *, and ]a il
]
- b. At least one of the.two source paths from the refueling water storage tank to the charging pump suction; and, '
?
- c. The flow path f rom the charging pump discharge to the Reactor H Coolant System via the regenerative heat exchanger.
APPLICABILITY: MODES.1, 2, 3, and 4. ACTION: 1 1
- a. With no boration source path from a boric acid storage tank OPERABLE,. -!
( ,
- 1. Demonstrate the OPERABILITY of the second source path from the f refueling water storage tank to the charging pump suction by i
verifying the flow path valve alignment; and ]
- 2. Restore the boration source path from a boric acid storage tank E to OPERABLE status within 72 hours or be in at least HOT STANDBY i and borated to a SHUTDOWN MARGIN equivalent to at least 1% ak/k at 200 F within the next 6 hours; restore the boration source jq path from a boric acid storage tank to ERABLE. status within :)
the next 72 hours or be in COLD' SHUTDOWN within the next 30. l hours, a
- b. With only one boration source path OPERABLE or the regenerative heat ]
exchanger flow path to the RCS inoperable, restore the required flow l paths to OPERABLE-status within 72 hours or be in at least HOT . ~ STANDBY and borated to a SHUTDOWN MARGIN equivalent to at le'ast 1% i ok/k at 200 F within the next 6 hours; restore at least two boration source paths to OPERABLE status within the next 72 hours or be in COLD SHUTDOWN within the next 30 hours.
- c. With the boration source path from a boric acid storage tank and the
. charging. pump discharge path via the regenerative heat exchanger-inoperable, within one hour initiate boration to a SHUTDOWN MARGIN ) equivalent to 1% Ak/k at 200 F and go to COLD SHUTDOWN as soon as l possible within the limitations of the boration and pressurizer level control functions of the CVCS.
*The flow required in Specification 3.1.2.2.a above shall be isolated from the other unit.
TURKEY POINT - UNITS 3 & 4 3/4 1-9 AMENDMENT N05.137AND 132
]
re. - - -
\
b REACTIVITY CONTROL SYSTEMS { b i
; SURVEILLANCE REQUIREMENTS :
-( ,
4.1.2.2 - The above required flow paths shall be demonstrated OPERABLE: At least.once per 7 days by verifying that the temperature of the-l- a. heat' traced portion of the flow path from the boric acid tanks is.
, greater' than or ' equal to 145 F when it is a required water. source; '
- b. cat least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, !
sealed, or otherwise secured in position, is in its correct position;
- c. At least once per 18 months by verifying that the flow path' required ;
by Specification 3.1.2.2a. and c. delivers at least 4 gpm to'the RCS. u 1 l l TURKEY POINT - UNITS 3 & 4 3/4 1-10 AMENDMENT N05.137AND 132
y , p REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION r 3.1.2.3 At least two charging pumps.with independent power supplies shall.be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. .With two charging pumps OPERABLE.and powered from a common power supply, restore at least two charging pumps from indeperndent power i supplies to OPERABLE status within 7 days or be in at least HOT I STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least-1% ok/k at 200 F within the next 6 hours; restore at least two 1 charging pumps from independent power supplies to OPERABLE status j within 72 hours or be in COLD. SHUTDOWN within the next 30 hours, i I
- b. .With only one charging pump OPERABLE, restore any two charging pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% ok/k at 200 F-within 6 hours; restore any two charging pumps to OPERABLE status within 72 hours or be in COLD .SHUiDOWN wit.hin the next 30 hours.
I
- c. The provisions of Specification 3.0.4 aru not applicable to ACTION a, f provided.the 7 day limit of ACTION a is not exceeded.
l i SURVEILLANCE REQUIREMENTS !
.i 4.1.2.3.1 The required charging pumps shall be demonstrated OPERABLE by testing pursuant to Specification 4.0.5 The provisions of Specification 4.0.4 ;
are not applicable for entry into MODES 3 and 4. i f lqh
.3 i
l TURKEY P0 INT - UNITS 3 & 4 3/4 1-11 AMENDMENT N05.137 AND 132 3
y i. [ REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE'- SHUT 00WN r LIMITING CONDITIN FOR OPERATION 3.1. 2. 4 As a minimum, one of the following borated water sources shall-be OPERABLE:
- a. A Boric Acid Storage System with:
- 1) A minimum indicated borated water volume of 500 gallons,
- 2) A boron concentration between 20,000 ppm and 22,500 ppm, and
- 3) A minimum solution temperature of 145oF. !
- b. The refueling water storage tank (RWST) with:
f i
- 1) A minimum indicated borated water volume of 20,000 gallons,
- 2) A minnum boron concentration of 1950 ppm,'and j
- 3) A minimum solution temperature of 39 F. ]
APPLICABILITY: MODES 5 and 6. ACTION: With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. SURVEILLANCE REQUIREMENTS 4.1.2.4 The above required borated water source shall be demonstrated OPERABLE:
- a. At least once per 7 days by: 4
- 1) Verifying the boron concentration of the water, .
- 2) Verifying the indicated borated water volume, and
- 3) Verifying the boric acid storage tank solution temperature when it is the source of borated water.
TURKEY POINT - UNITS 3 & 4 3/4 1-12 AMENDMENT N05.137AND 132 1
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,3 , ,REACTIVITYrCONTROL SYSTEMS '
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'M . LSURVEILLANCE REQUIREMENTS (Continued)=
~ 4
- g. -
a .. . . . - 1 V, A' t b'. .-_By2 verifying the RWST temaeraturesis'above.itsLlimit?whenever the 4
. outside air temperatureTis less than 39 F at the'fol. lowing . 'J
- c. . frequencies:: ', !
3
~
, - . :1):Within-one hour when the outside_: tempersture isibelow 39*F for_- j.
L23' consecutive hours.:and ?! t 1 . 2) AtEleast once perL24 hours when_the catside. temperature is'below. 4 39'E. . 1
.8
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, TURKEY POINT -
UNITS 3'& 4 AMENDMENT N05.-137AND 132 t
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REACTIVITY CONTROL SYSTEMS , q g S, BORATED WATER SOURCES:k OPERATIkGL < ) 1 -c LIMITING CONDITION FOR OPERATION p- l t , .n
. . _ s
'3.1.~2,5= The following borated water sources shall be OPERABLE:
(; -
- a. A Boric ^ Acid- Storage System with; Ej f, .
'j 1
P 1)'AminimumiNdicatedboratedwatervolumeof3080 gall'ons,-
- 2) A boron concentration between 20,000 ppm and:22,500 ppm, and -!
lf 1 , -
- 3) A minimum solution. temperatureof 145 F.
l 9 4 b. The refueling water storage tank (RWST) with: r J1); Aiminimum indicated borated water-volume of 320,000 gallons, j
- 2) A minimum boron concentration.of 1950 ppm,-
i
- 3) LA minimum solution temperature of 39 F,'and }
, 4) A maximum solution temperature of 100 F. .
I APPLICABILITY: MODES 1, 2, 3, and 4. .
' ACTION:
- a. _W ith the required Boric Acid Storage System inoperable verify- that-the-RWST is 0PERABLE; restore the system to OPERABLE status within j 72 hours 1or be in-at least HOT. STANDBY.within the'next 6. hours.and - 1
-borated'to--a SHUTDOWN: MARGIN equivalent to at least 1%..Ak/k at 200 F;_ 4 restore the-Boric Acid Storage System to OPERABLE status within the ;
next 72 hours or be in COLD'SHVTDOWN within the next 30 hours. , L
- b. -With the-RWST inoperable, restore the tank to OPERABLE status
~
s within 1 hour or be in at least HOT STANDBY within the next: l 6 hours and in COLD SHUTDOWN within the following 30 hours. 4 q .p I a: t i
-TURKEY-POINT - UNITS-3 & 4 3/4 1-14 AMENDMENT N05.137A ND 132 l
k g 'l .
5 i P REACTIVITY CONTROL SYSTEMS
- SURVEILLANCE REQUIREMENTS-4.1.2.5 Each borated water source shall be demonstrated OPERABLEi
- a. At'le ' once per 7 days vy.
. 1) Verifying the boron concentration in the water, 2)' Verifying the indicated borated water volume of the water-source, and
- 3) Verifying the Boric Acid Storage System solution temperature when it is the source of borated water,
- b. By verifying the-RWST temperatur's is within limits whenever the outside air temperature is less than 39 F or greater than 100 F at the following frequencies:
- 1) Within one hour upon the outside temperature exceeding its limit for 23 consecutive hours, and
- 2) At 1 east once per 24 hours _while the outside temperature exceeds its limits.
TURKEY POINT - UNITS 3 & 4 3/4 1-15 AMENDMENT N05.137AND 132
3 e e [
' REACTIVITY CONTROL SYSTEMS HEAT TRACING LIMITING CONDITION FOR OPERATION 3.1.2.6 At least two independent channels of heat tracing shall be OPERAB'LE for the boric acid storage tank and for the heat traced portions of the-associated flow paths required by Specification 3.1.2.2.
APPLICABILITY: MODES 1, 2, 3 and 4 MODES-5 and 6 (when the boric acid storage tank is the borated water source per Specification 3.1.2.4) ACTION: MODES 1, 2, 3 a'nd 4 With only one chtnnelcof heat tracing on either the boric acid storage tank or on the heat traced portion of an associated flow path OPERABLE, operation may continue for up to 30 days providedtthe-tank and flow path temperatures are verified to be greater than or equal to 145 F at least once per 8 hours; otherwise, be in-at least HOT STANDBY within 6 hours and-in COLD SHUTDOWN within the fol110 wing 30 hours. MODES 5 and 6 With only one channel of heat tracing on either the boric acid storage tank or on the heat traced portion of an associated flow path OPERABLE, operations involving CORE ~ ALTERATIONS'or positive reactivity additions.may continue for up to 30 days provided the tank and_ flow path temperatures are verified to be greater than or equal to 145 F at least once per 8 hours; otherwise, suspend all activities involving CORE ALTERATIONS or positive reactivity changes.
~
SURVEILLANCE REQUIREMENTS ; 4.1.2.6 Each heat tracing channel for the boric acid storage tank and associated flow path required by Specification 3.1.2.2 shall be demonstrated OPERABLE:
- a. At least once per 31 days by energizing each heat tracing channel, and
- b. At least once per 7 days by verifying the tank and flow path temperatures to be greater than or equal to 145 F. The tank temperature shall be determined by measurement. The flow path temperature shall be determined by either measurement or recirculation flow until establishment of equilibrium temperatures within the tank.
TURKEY PolNT - UNITS 3 & 4 3/4 1-16 AMENDMENT N05.137AND 132
r; , , a f 9EACTIVITY CONTROL SYSTEMS L 3/4.1.3 MOVABLE' CONTROL ASSEMBLIES GROUP HEIGHT , LIMITING CONDITION FOR OPERATION' 3.1.3.1 All full length-(shutdown and ccatrol) rods shall be OPERABLE and l positioned within i:12 steps (Analog Rod Position Indication) _of the group step - l counter demand position within one hour af ter rod motion. APPLICABILITY: NODES,1* and 2* ACTION:'
- a. With one or more full length rods inoperable due to being immovable as a result of excessive friction or mechanical interference or known-to be unti'ippable, determine that the SHLTTDOWN MARGIN requiremer,t of I
Specification 3.1.1.1 is satisfied within 1 hour and be in HOT STANDBY within 6 hours.
- b. With more than one full length rod inoperable or misaligned from the group step counter demand position by more.than 12 steps (Analog Rod Position Indication), be in HOT STANDBY within 6 hours.
- c. With one full length rod inoperable due to causes other than :
addressed by ACTION a, above, or misaligned from its group step counter demand position by more than i 12 steps (Analog Rod Position Indication), POWER.0PERATION may continue provided that within one hour either:
- 1. The rod is restored to OPERABLE status within the above alignment l requirements, or
- 2. The remainder of the rods in the bank with- the inoperable rod are '
aligned to within 12 steps of the inoperable rod while maintaining the rod sequence and insertion limits of-Figure 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or
- 3. The rod is declared inoperable and the SHUIDOWN MARGIN '
. requirement'of Specification 3.1.1.1 is satisfied. POWER OPERATION may then continue provided that:
l l l l
^5ee Special Test Exceptions 3.10.2 and 3.10.3.
' TURKEY POINT - UNITS 3 & 4 3/4 1-17 AMENDMENT N05.137AND 132 1
w m
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?
REACTIVITY CONTROL' SYSTEMS m f LIMITING CONDITION FOR OPERATION (Continued) b<,'* E a) The THERMAL POWER level is. reduced tofless .than or equal to 75%'of.
' RATED THERMAL POWER within one hour and'within the next 4 hours the e
9
. power range neutron flux high trip setpoint~is' reduced to less than-
~
. , or equal to 85% of RATED THERMAL' POWER.- THERMAL' POWER shall be maintained less than or equal to 75% of RATED THERMAL POWER until F '
compliance with ACTIONS 3.1.3.1.c.3.c and 3.1.3.1.c.3.d below.are demonstrated, and. q b) The SHUTDOWN MARGIN requirement of Specification 3.1,1.1 is
, ' determined at least once per 12 hours, and.
c)~ A power distribution map is obtained.from the movable-incore N detectors and F9(Z) and F 3H are verified to be within their limits within 72 hours, and d) A reevaluation of each eccident analysis of Table 3.1-1 is performed n within 5 days; this reevaluation shall-confirm that the previously , d" analyzed results of these accidents remain valid for the duration of H operation under these conditions. J J 5URVIELLANCE REQUIREMENTS _ 4
.4.1.3.1.1 The position of each full length-rod shall be determined to be
.within i 12 steps (Analog Rod Position Indication).of the group step counter demand position at least once per-12 hours,(allowing for one hour thermal soak after rod motion) except during. time'invervals when the Rod Position Deviation -
1 Monitor is inoperable, then verify the group positions at least once per , 4 hours. ' 4.1.3.1.2 Each full length rod not fully inserted in the core shall be deter-mined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days. i i t 132 TURKEY POINT - UNITS 3 & 4 3/4 1-18 AMENDMENT N05.137AND
- i .. -
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. TABLE 3; 1 3 1
e ACCIDENT A'NALYSES REQUIRING REEVALUATION: . i SIN THE EVENT OF AN INOPERABLE FULL-LENGTH ROD-.. Rod Cluster Contro1' Assembly Insertion Characteristics-
~[
J
. Rod Cluster Control Assembly Misalignment ,
i
< o , Loss of Reactor Coolant from Small: Ruptured Pipes or from Cracks in= , j
'Large Pipes.Which Actuates the Emergency Core Cooling System- 1 b Single' Rod Cluster Control, Assembly Withdrawal at Full Nivr:1 .
j;
~ .
.MajorReactorCoolantSystemPipeRuptures(Loss-of.-Coolant-
-Accident)-
. Major Secondary-. Coolant. System Pipe Rupture l Rupture'of a' Control' Rod Drive' Mechanism Housing (Rod Cluster Control e ^ Assembly Ejection).
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(TURKEY POINT . UNITS 3 & 4 3/4 1-19 AMENDMENT N05 137AND 132 l, , l
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y , , [> , , , 4 %9 , LREACTIVITY-CONTR06 SYSTEMS' ys ,' 1 W - > POSITION-INDICATION SYSTEMS'- OPERATING
.. j
+
" LIMITING CONDITION FOR'0PERATION i z u
, , 13.1.3.2 The Analog Rod Position Indication' System an'd the Demand 1 Position -
Indication' System shall be OPERABLE and capable of determining the'respectiv'ei -l actual and demanded shutdown and. control rod positions as:follows:' ~
- a. Analog rod positionDindicators,-within one-hour after rod.motioni ,
(allowance for thermal soak); , 5 r E.
. All Shutdown Banks: '1 12 steps:of the group demand couaters for-withdrawal ranges of 0-30-steps and 200-228 steps, t
. Control BanklA and B: 1 12. steps of the group demand counters for. i withdrawal ranges of 0-30 steps and 200-228 steps.
Control Banks C and D: 1-12 stepsfof the group demand countersjfor
- withdrawal range of=0-228 steps, b., Group demand counters; 2 steps. .,
a
'7 APPLICABILITYi MODES 1 and 2.
-ACTION: '
- a. With a" maximum of orie ,ai.2 lag rod position indicator- per bank inoperable' 1 either: 4 N 1. Determinetheposition.-of~the.non-indicating _ rod (s) indirectly 3 1 by the: movable incore detectors ,at least once per 8 hours and ' l y
, within one' hour.af ter any. motion- of the non-indicating? rod-which
_ . exceeds-24 steps -in one direction since the last determination
, of the rod's position, or j n b
- 2. Reduce THERMAL POWER to less than 75% of RATED THERMAL POWER-within 8 hours,
- b. -With a. maximum of one demand position indicator per bank inoperable--
either:
- E Verify.that all analog rod position in'dicators for the affected bank are OPERABLE and that the most withdrawn rod and the least- !
# withdrawn rod of the bank are within a maximum of 12 steps.of.
~
each other at least once per 8 hours, or ,. a
- 2. Reduce-THERMAL POWER to less than 75% of RATED THERMAL POWER 4, w'ithin 8 hours. ;
4
. TURKEY POINT - UNITS 3 & 4 3/4'l-20 AMENDMENT N05.137AND 132 o
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-4.1.3!2.1? Each analog-rod position ~ indicator shall_be determined to be 0PERABLE-by verifying'that the Demand Position'Indi'ation c System and the Analog Rod Posi- !
- tion, Indication System ~ agree =within 12 steps.(allowing for one hour thermal < soak after rod-motion) attleast once per 12 hours except during time-intervals when .
the Rod Position Deviation Monitor is inoperable,- then compare'the Demand Posi ~ 'i tion Indicati_on System and the Analog Rod Position Indication System at least once per 4' hours.
~
( 4.1.3.2.2 Each of the above required analog rod-position-indicator (s) shall be ' f.L - determi'ned to'be OPERABLE by performance of a CHANNEL CHECK,-' CHANNEL CALIBRA-- j 3~ TION and ANALOG CHANNEL OPERATIONAL TEST performed in accordance with ' cTable 4.1-1-. ~ - M l 1 i r 5 1
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. TURKEY P91NT - UNITS 3 & 4 3/4 1-21 AMENDMENT N05.137 AND 132.
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c.* iROD POSITION" INDICATOR' SURVEILLANCE REQUIREMENTSi . pb', ; >;
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- :TURKEYJPOINT - UNITS ,3 & 4' 3/4 1-22 AMENDMENT N05)37 AND 132 ;
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L-4e ._ J. REACTIVITY-CONTROL' SYSTEMS POSITION INDICATION-SYSTEM - SHUTDOWN' , LIMITING CONDITION FOR OPERATION
;3.1.3.3 The group step counter demand position indicator shall be OPERABLE-
-and capable of determining within i 2 steps'the demand position for each shut-down and control rod not f ully inserted.
APPLICABILITY: MODES 3* **, 4* **, and 5* ** ACTION: With less than the above required group step counter demand position indica-tor (:) OPERABLE, open the reactor-trip system breakers,
- SURVEILLANCE REQUIREMENTS
.i
-4.1;3.3.1 Each of the above required group step counter demand position,indi- l cator(s) shall~be determined to be OPERABLE by' movement of the associated- ~4 control rod at least 10 steps in any one direction at least once per 31 days,
~4.1.3.3.2 A CHANNEL CHECK CALIPRATION AND ANALOG CHANNEL OPERATIONAL TEST shall be performed'per Table 4.1-1. t f
\
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*With the Reactor Trip System breakers in the closed position.
**See Special Test Exceptions Specification 3.10.4.
TURKEY POINT - UNITS 3 & 4 3/4 1-23 AMENDMENT N05.137AND 132 l
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' REACTIVITY CONTROL SYSTEMS
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ROD DROP' TIME-k , LIMITING CONDITION FOR-OPERATION < e L - 3.1.3.4 The individual fuli-length-(shutdown and control) rod drop' time from the fully withdrawn position shall ce less than or_ equal'to 2.4',econds from beginning of decay of stationary gripper coil. voltage to dashpot entry with: ,
- a. 7,yg greater than or' equal to 541'F, and j[, b. ' AllLreactor coolant pumps operating.
" APPLICABILITY: MODES 1 and 2.
he ACT10N:' With the drop time of any.fulbhingth rod determined to exceed the above l limit,- restore the rod drop time k withh. the above limit prior, to proceeding- i p , to MODE 1 or 2. [ . 4 i-r . SURVE]LLANCE REQUIREMENTS p F - 4.1.3.4' The rod drop time of full-length rods shall be demonstrated through
' measurement prior to reactor et(ticality:
- a. For all rods follwirig e3ch removal of the reactor vessel head, V ,
L , b. - For specifically r?fected individual rods following-any maintenance h on or modificdion to the Control Rod Drive System which could ,; affect the drop time of those specific rods, and
- c. At least once.per 18 months.
n i( ( b,
; TURKEY POINT - UNITS 3 & 4 3/4 1-24 AMENDMENT N05.137AND 132 3 L
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<SHUIDOWN ROD INSERTION LIMIT o r !
; LIMITING CONDITION FOR OPERATION 4
. 3.1.3.5 All shutdown rods shall be fully withdrawn. R APPLICABILITY: MODES 1* and 2^ ** '
1 .: i k ACTION:- , 1 M With'a maximum of one shutdown rod not fully withdrawn, except for surveillance l y* w
. testing pursuant to Specification 4.1.3.1.2, within I hotr either: i n
L j a. Fully withdraw the rod, or . l m ,. 1 M b. Declare the rod ~to be-inoperable and' apply Specification i i, 3.1.3.1. .! i
]:;
,; SURVEILLANCE REQUIREMENTS ..1 t
?
4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn:. .,
- a. Within 15 minutes prior to withdrawal of any rods in contro1L banks A.,B, C, or D during an approach toireactor criticality, and j
.:b. 'At least once per 12 hours thereafter, a
r i
)
*See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
**With K,ff greater than or equal to 1.0 e- oTURKEY POINT ' UNITS 3 & 4 3/4 1-25 AMENDMENT N05.137AND 132 ;
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$ CONTROL: ROD INSERTION LIMITS '
- v,
- LIMITING CONDITION FOR OPERATION a '3.1.3.6 :The: control banks shall be limited in physical insertion as shown in' Figure 3.1-2.
r
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APPLICABILITY: MODES 1* and 2* ** 1 ACTION: y g With the control banks; inserted beyond the above insertion limits,'except for surveillance testing pursuant to Specification 4.1.3.1.2 either:. ( [ , _
- a. Restor'e the control banks to within the-limits within 2 h'ours, or
- b. Reduce THERMAL POWER within two hours to'less than.or equal to that; fractinn of RATED ~ THERMAL POWER which is allowed by the~ bank posi-tion using the above figures, or. ,
c .' - Be in at~1 east HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS L ;; . i 4.I'.3.6 The position of frach control bank shall be determined to be within
'the insertion limits'at Teast~once per 12. hours, except during time. intervals 1 when the Rod Insertion Limit Monitor is inoperable. then verify.the individual-rod positions at least onte per-4 hours, li 7
o e' i l i L
*See Special Test Exceptions Specifications 3.10.2 and 3.10.3.
**With.Kgf( greater than or equal to 1.0 i f
, TURKEY POINT - UNITS'3 & 4 3/4-1-26 AMENDMENT N05.137AND 132
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- RATED THERMAL POWER flGURE 3.1-2 pophghUfg M W % ER HR g TURKEY POINT - UNITS 3 & 4 3/4 1-27 AMENOMENT NOS.137AND 132
l 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL i :.'X DIFFERENCE (AF's) shall be maintained within a't 5% target band (flux d'rference units) about the target flux difference. The indicated AFD may deviate outside the above required target band at greater than or equal to 50% but less than 90% of RATED THERMAL POWER provided the indi-cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumu-
. lative penalty deviation time does not exceed I hour during the previous 24 hours.
The indicated AFD may deviate outside the above required target band at greater than 15% but less than 50% of RATED THERMAL POWER provided the cumulative penalty deviation time does not exceed I hour during the previous 24 hours. APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER.* ACTION:
- a. With the indicated AFD outside of the above required target band and with THERMAL POWER greater than or equal to 90% of RATED THERMAL POWER, within 15 minutes either: .
- 1. Restore the indicated AFD to within the target band limits, or
- 2. Reduce THERMAL P0.,cR to less than 90% of RATED THERMAL POWER.
, b. With the indicated AFD outside of the above required target band for l- more than 1 hour of cumulative penalty deviation time during the previous 24 hours or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less than 90% but equal to or greater than 50% of RATED THERMAL POWER:
- 1. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER -
within 30 minutes, and
- 2. Reduce the Power Range Neutron Flux * ** - High Trip Setpoi ds to I less than or equal to 55% of RATED THERMAL POWER within tne next 4 hours.
l c. With the indicated AFD outside of the above required target band for i more than 1 hour of cumulative penalty deviaticn time during'the l previous 24 hours and with THERMAL POWER less than 50% but greater 1 l
~
l
*See Special Test Exceptions Specification 3.10.2.
l
** Surveillance testing of the Po- %nge Neutron Flux Channels may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the Acceptable Operation Limits of Figure 3.2-1. A total of 16 hours operation may be accumulated with the AFD outside of the above required target band during testing without penalty deviation.
TURKEY POINT - UNITS 3 & 4 3/4 2-1 AMENDMENT N05.137AND 132 I
4 POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) ACTION (Continued) than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is witnin the above required target band.
. SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:
- a. Monitoring the indicated AFD for each OPERABLE excore channel:
- 1) At least once per 7 days when the alarm used to monitor the AFD is OPERABLE, and
- 2) At least once per hour for the first 6 hours after restoring the alarm used to monitor the AFD to OPERABLE status.*
- b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the alarm used to monitor the AFD is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.
4.2.1.2 The indicated AFD shall be considered outside of its target band when two or more OPERABLE excore channels are indicating the AFD to be outside the
- target band. Penalty deviation outside of the above required target band shall l- be accumulated on a time basis of
- a. ' One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and
- b. One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.
4.2.1.3 The target flux difference of each OPERABLE excore channel shall be i determined by measurement at least once per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable. 4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and the predicted value at the end of the cycle ~ life. The provisions of Specification 4.0.4 are not applicable.
" Performance of a functional test to demonstrate OPERABILITY of the alarm used to monitor the AFD may be substituted for this requirement.
TURKEY POINT - UNITS 3 & 4 3/4 2-2 AMENDMENT N05.137AND132 ,
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l- - 60 - 40 - 30 - 20 - 10 0 10 20 30 40 50 - Fl.UX DIFFERENCE (Alp 1 i
~
FIGURE 3.2-1 't AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER TURKEY POINT - UNITS 3 & 4 3/4 2-3 AMENDMGT NOS.137 AHD 132
- T
,m_-m _ _ , , . _ _ , _ _ , . _ . , , , , . . _ . . _. m_,_ , . - ,
l s ; POWER DISTRIBUTION LIMITS .l l 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F0 (Z) LIMITING CONDITION FOR OPERATION
3.2.2 Fh(Z)shallbelimitedbythefollowingrelationships
X [K(Z)) for P > 0.5 F (Z) 1 +[F )l l L F (Z) 1 X [K(Z)] for P 1 0.5 where: [Fq )l = 2.32 limit P= Thermal Power , Rated Thermal Power J [F ]M = The Measured Value, and 0 , K(Z) is the function obtained from Figure 3.2-2 for a given core height location. APPLICABILITY: MODE 1 + ACTION: With the ' measured value of F (Z) exceeding its limit: a., > ReduceTHERMALPOWERatleast1%foreach1%F(Z)exceedsFh within 15 minutes and similarly reduce the Power Range Neutron Flux - High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION n:ay : l' proceed provided the Overpower Delta-T Trip Setpoints (value of K ) ; and. ; havebeenreducedatleast1%foreach1%F(Z)exceedstheFh(Z) l b. Identify and correct the cause of the out-of-limit condition prior r I to increasing THERMAL POWER above the reduced power limit required l by ACTION a , above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit. L 1 l TURKEY POINT - UNITS 3 & 4 3/4 2-4 AMENDMENT NOS.137AND132 e i
-- =-
n - , - . . . . _ . . . . ,
i l l 1 (6.0,1.0) 3.o ______________ 22= ~ _ .y.sgs4L ; c.s 3( ' 04
-\ -
S \ o.7
"* \
(12.0.0.647) , M 0.5 l ( @ 0.4 ! o.3 0.2 0.1 i ! 0.0 0 1 2 3 4 5 6 7 8 s 10 11 12 CORE HEIGHT (FT.) 1 1 FIGURE 3.2-2 K(Z) NORMALIZED Fq(Z) AS A FUNCTION OF CORE HEIGHT TURKEY POINT - UNITS 3 & 4 3/4 2-5 AMENDMENT N05.137 AND132
POWER DISTRIBUTION LIMITS , SURVEILLANCE REQUIREMENTS 4.2.2.1 If (F ]P q as predicted by approved physics calculations is greater than (Fq )l and P is greater than P
- as defined in 4.2.2.2, F (Z) shall be T 9 evaluated by MIDS (Specification 4.2.2.2), BASE LOAD (Specification 4.2.2.3) or RADIAL BURNDOWN (Specification 4.2.2.4) to determine if F q is within its limit [F ]P q = Predicted Fq ). ,
If (F g]P, is less than [F q)l or P is less than P , F (Z) shall be evaluated to T 9 determine if F q (Z) is within its limit as follows: ,
- a. Using the movable incore detectors to obtain power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER. '
- b. Increasing the measured F (Z) component of the power distribution 9
map by 3%'to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. ' Verifying that the requirements of Specification 3.2.2 are satisfied. c. F(2)5Fh(2) Where Fl(Z) is the measured9 F (Z) increased by the allowance for manu-facturing tolerances and measurement uncertainty and lF (Z) is the F limit defined in 3.2.2. 0 9
- d. Measuring F (Z) according to the following schedule:
- 1. Prior to exceeding 75% of RATED THERMAL POWER,** after refueling,
- 2. At least once per 31 Effective Full Power Days,
- e. With the relationship specified in Specification 4.2.2.1.c above not being satisfied:
- 1) Calculate the percent F (Z) exceeds its limit by the following expression:
l FhZ) -1 X 100 for P > 0.5 [Fq ]' X K(Z)/P _
=
F (Z) -1 X'100 for P < 0.5 . [Fq ]l X K(Z)/0.5 , j
*Py = Reactor power level at which predicted Fq would exceed St.s limit. **During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and power distribution map obtained.
TURKEY POINT - UNITS 3 & 4 3/4 2-6 AMENDMENT N05 137AND 132 e e
f, I. POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) ;
- 2) The following action shall be taken:
]
a) Comply with the requirements of Specification 3.2.2 for F (Z) exceeding its limit by the percent calculated above. < 4.2.2.2 MIDS \ GP: ationis pe mitted at power above P where PT equals the ratio of [F )l T q divided oy [Fq) if the following Augmented Surveillance (Movable Incore j Detection System,' MIDS) requirements are satisfied: ' i
- a. The axial power distribution shall be measured by MIDS when required ]
such that the limit of [F )lq /P times Figure 3.2.2 is not exceeded. Fj (Z) is the normalized axial power distribution from thimble j at core elevation (Z).
- 1. If F j(Z) exceeds [F j(Z)]s* as defined in the bases by 1 4%,
immediately reduce thermal power one percent for every percent by which [Fj (Z)]s is exceeded.
- 2. j exceeds [F (Z)]s If F (Z) j by > 4% immediately reduce, thermal ,
power below P . Corrective action to reduce Fj(Z) below the T limit will permit return to thermal power not to exceed current P ** as defined in the bases, L
- b. Fj (Z) shall be determined to be within limits by using MIDS to monitor the thimbles required rier Specification 4.2.2.2.c at the '
following frequencies.
- 1. At least once every 24 hours, and
- 2. Immediately following and as a minimum at 2, 4 and 8 hours following the events listed below and every 24 hours thereafter.
- 1) Raising the thermal power above P T
, or
- 2) Movement of control-bank D more than an accumulated total of 15 ste)s in any one direction.
- c. MIDS shall be opera)1e when the thermal power exceeds TP with:
. 1.
At least in the twohave bases thimbles been determined. available for which it3 and d as defined '
*[Fj(Z)]3 is the alarm setpoint for MIDS.
**P is reactor thermal power expressed as a fraction of the Rated Thermal Pbwerthatisusedtocalculate[F(Z)]s' j TURKEY POINT - UNITS 3 & 4 3/4 2-7 IMENDMENT N05.137AND 132
~
1 l. i ; l l POWER DISTRIBUTION LIMITS ) L SURVEILLANCE REQUIREMENTS (Continued) i p 2. At least two movable detectors available for mapping Fj (Z).
- 3. The continued accuracy and representativeness of the selected thimbles shall be verified by using the most recent flux map to updatetheEforeachselectedthimble. The flux map must be !
updated at least once per 31 effective full power days. I where: R = Total peaking factor from a full flux map ratioed to the axial peaking factor in a selected thimble. 1 j - The thimble location selected for monitoring, i 1 4.2.2.3 Base Load Base. Load operation is permitted at powers above T P if the following require-ments are satisfied:
- a. Either of the following preconditions for Base Load operation must !
be satisfied, l
- 1. For entering Base Load operation with power less than PT '
i a) Maintain THERMAL POWER between P /1.05 and P f r at least l T T , 24 hours, l b) Maintain the AFD (Delta-1) to within a i 2% or i 3% 1 target band for at least 23 hours per 24-hour period. i c) After 24 hours have elapsed, take a full core flux map to 1 determine F (Z) unless a valid full core flux . nap was taken within the time period specified in 4.2.2.1d. d) . Calculate PBL per 4.2.2.3b.
- 2. For entering Base load operation with power greater than PT '
l-a) Maintain THERMAL POWER between P and the power limit - determinedin4.2.2.2foratlealt24 hours,andmaintain Augmented Surveillance requirements of 4.2.2.2 during this l period. - 1 b) Maintain the AFD (Delta-I) to within a i 2% or i 3% target band for at least 23 hours per 24-hour period, I TURKEY POINT - UNITS 3 & 4 3/4 2-8 AMENDMENT N05.137AND 132 1 _ i - .
r POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) c) After 24 hours have elapsed, take a full core flux map to ' determine F (Z) unless a valid full core flux map was taken within the time period specified in 4.2.2.1d. d) Calculate P BL Pe r 4. 2. 2. 3b.
- b. Base Load operation is' permitted provided:
- 1. THERMAL POWER is maintained between P and P BL r between PT and 100% (whichever is mot limiting)T
- 2. AFD (Delta-1) !s maintained within a i 2% or i 3% target band.
- 3. Full core flux maps are taken at least once per 31 effective Full Power Days.
P BL and PT are defined as: . P BL
- If 0]l X K(Z)
Fl(Z)XW(Z)BLX1.09 P T
= [Fq/q
)l [F ]P where: F (Z) is the measured Fq (Z) with no allowance for manufactur-ingtolerancesogmeasurementuncertainty. For the purpose of this l
'i Specification [FQ(Z)] shall be obtained between elevations bounded by 10% and 90% of the active core height. [Fq ) is the Fq limit.
K(Z) is given in Figure 3.2-2. W(Z) BL is the cycle dependent function that accounts for limited power distribution transients ' encountered during base load operation. The function is given in the Peaking Factor Limit Report as per 7 Specification 6.9.1.6. The 9% uncertainty factor accounts for manufacturing tolerance, measurement error, rod bow and any burnup and power dependent peaking factor increases,
- c. During Base Load operation, if the THERMAL POWER is decreased below PT , then the conditions of 4.2.2.3.a shall be satisfied before re-entering Base Load operation.
- d. If any of the conditions of 4.2.2.3b are not maintained, reduce THERMAL POWER to less than or equal to P ,
T r, within 15 minutes ! initiate the Augmented Surveillance (MIDS) requirements of 4.2.2.2. TURKEY POINT - UNITS 3 & 4 3/4 2-9 AMENDMENT NOS.137 AND 132
F POWER DISTRIBUTION LIMITS L , SURVEILLANCE REQUIREMENTS (Continued)
$ 4.2.2.4 RADIAL BURN 00WN Operation is permitted at powers above P T if the following Radial Burndown conditions are satisfied:
- a. Radial Burndown operation is restricted to use at powers between P T
- and PRB or PT and 1.00 (whichever is most limiting). The maximum relative power permitted under Radial Burndown operation, PRB, is equaltotheminimumvalueoftheratioof[Fh(Z)]/[F(Z)]RB Meas, 9
where: [Fg(Z))RB Meas. = [Fxy(2)3 Map Meas. x F,(Z) x 1.09 and [Fh(Z)]isequalto(Fh]xK(Z).
- b. A Tull core flux map to determine (Fxy(I)3M&p Heas shall be taken within the time period specified in Section 4.2.2.1d.2. For the pur-pose of the specification, [F x Meas, shall be obtained between the elevations boundedand by90%10% y(Z))g,p of the active core Leight.
- c. The function F2 (Z), provided in the Peaking Factor Limit Report l
(6.9.1.6), is determined analytically and accounts for the most per-turbed axial power shapes which can occur under axial power distribu-
- l. tion control. The uncertainty factor of 9% accounts for manufacturing !
tolerances, measurement error, rod bow, and any burnup dependent peak- i ing factor increases. ! d .' Radial Burndown operation may be utilized at powers between P and T PRB, or, PT and 1.00 (whichever is most limiting) provided that the l AFD (Delta-I) is within i 5% of the target axial offset. l l' e. If the requirements of Section 4.2.2.4d are not maintained, then the power shall be reduced to less than or equal to P T
, or within L 15 minutes Augmented Surveillance of hot channel factors shall be initiated if the power is above PT '
o 4.2.2.5 When F (Z) is measured for reasons other than meeting the requirements 9 i of Specifications 4.2.2.1, 4.2.2.2, 4.2.2.3 or 4.2.2.4 an overall measured F (Z) shall be obtained from a power distribution map and increased by 3% to 9 account for manufacttiring tolerances and further increased by 5% to account for - measurement uncertainty. TURKEY POINT - UNITS 3 & 4 3/4 2-10 AMENDMENT NOS.137AND 132
i I POWER DISTRIBUTION LIMITS - 1 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR l LIMITING CONDITION FOR OPERATION 3.2.3 Ffg shall be limited by the following relationship: ) FfH < 1.62 [1.0 + 0.3 (1-P)], 1 Where:- THERMAL POWER j p _ RATED THERMAL POWER APPLICABILITY: MODE 1. I ACTION: With FfH exceeding its -limit:
- a. Within 2 heurs either:
- 1. RestoreF[htowithintheabovelimit,or
- 2. Reduce THERMA; POWER to less than 50% of RATED THERMAL POWER' -
'and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the .
next 4 hours. I
- b. Within 24 hours of initially being outside the above limit, verify
' through incore flux mapping that Ffg has been restored to within the above limit, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours,
- c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit -
required by ACTION a.2. and/or b., above; subsequent POWER OPERATION may proceed provided that F H is demonstrated, through incore flux ; mapping, to be within the limit of acceptable operation prior to . exceeding the following THERMAL POWER levels:
- 1. A nominal 50% of RATED THERMAL POWER, -
- 2. A nominal 75% of RATED THERMAL POWER, ar.d ,,
- 3. Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER.
TURKEY POINT - UNITS 3 & 4 3/4 2-11 AMENDMENT N05.137AND 132 4 i
I POWER DISTRIBUTION LIMITS i i l SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. ! 4.2.3.2 WhenameasurementofFfH istaken,themeasuredFfg shall be increased by 4% to account for measurement error. , 4.2.3.3 ThiscorrectedFfg shall be determined to be within its limit through incore flux mapping:
- a. Prior to operation above 75% of RATED THERMAL POWER after each fuel ,
loading, and '
- b. At least once per 31 Effective Full Power Days.
l i 1
)
)
O t l l I l
+
3 l TURKEY POINT - UNITS 3 & 4 3/4 2-12 AMENDMENT NOS.137AND 132 7
POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER *. ACTION:
- a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:
a) The QUADRANT POWER TILT RATIO is re'duced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
- 2. Within 2 hours either:
a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at.least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron
- Flux-High Trip Setpoints within the next 4 hours, e
- 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL i
POWER to less than 50% of RATED THERMAL POWER within the next l 2 hours and reduce the power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and l 4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL F0WER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER. -
*See-Special Test Exceptions Specification 3.10.2.
TURKEY POINT - UNITS 3 & 4 3/4 2-13 AMENDMENT N05.137 AND 132
POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued) ACTION (Continued)
- b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either: :
I a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or : 1 b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER.
- 2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for i each 1% of indicated QUADRANT POWER TILT RATIO in excess of i 1, within 30 minutes; ,
- 3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours after exceeding the limit or reduce THERMAL i POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip ,
'Setpoints to less than or equal to 55% of RATED THIRMAL POWER within the next 4 hours; and t
- 4. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER;. subsequent POWER OPERATION
- above 50% of RATED THERMAL POWER may proceed provided that the -
QUADRANT POWER TILT RATIO is verified within its limit at least. > once'per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.
- c. With sne QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control rod:
- 1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either: ' .
a) The QUADRANT POWER TILT RATIO is reduced to with its limit, or 1 . b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. l
't TURKEY POINT - UNITS 3 & 4 3/4 2-14 AMENDMENT N05.137AND 132 L ,
)
l POWER DISTRIBUTION LIMITS l LIMITING CONDITION FOR OPERATION (Continued) 'l ACTION (Continued)
- 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High ;
Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and
- 3. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the )
QUADRANT POWER TILT RATIO is verified within its limit at least . once per hour for 12 hours or until verified at 95% or greater
-RATED THERMAL POWER.
- d. The provisions of Specification 3.0.4 are not applicable. ]
1 SURVEILLANCE REQUIREMENTS 1 4.2.4.1 The QUADRAN1 POWER TILT RATIO shall be determined to be within the i limit above 50% of RATED THERMAL POWER by: l
- a. Calculating the ratio at least once per 7 days when the Power Range Upper Detector High Flux Deviation and Power Range Lower Detector ;
High Flux Deviation Alarms are OPERABLE, and ' l
- b. Calculating the ratio at least once per 12 hours during steady-state i operation when either alarm is inoperable. 1 4.2.4.2 The QUADRANT POWER TILT RATIO shall be determined to be within the +
limit when.above 75% of RATED THERMAL POWER with one Power Range channel
-inoperable by using the movable incore detectors to confirm that the normalized symmetric power distribution, obtained either from two sets of four symmetric 1 thimble locations or full-core flux map, or by incore thermocouple map is I consistent'with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours.
4.2.4.3 If the QUADRANT POWER TILT RATIO is not within its limit within 24 hours and the POWER DISTRIBUTION LIMITS of 3.2.2 and 3.2.3 are within their r limits, a Special Report in accordance with 6.9.2 shall be submitted within 30 days including an evaluation of the cause of the discrepancy. ' t I l i TURKEY POINT - UNITS 3 & 4 3/4 2-15 AMENDMENT N05.137AND 132 l l
~ .
1
POWER DISTRIBUTION LIMITS. 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION l 3.2.5 The following DNB-related parameters shall be maintained within the following limits. .
- a. Reactor Coolant System T avg 5 576.6*F b.- Pressurizer Pressure > 2209 psig*, and I
- c. Reactor Coolant System Flow >277,900 gpm APPLICABILITY: MODE 1.
~
' ACTION:
With any of the above' parameters exceeding its limit, restore the parameter to within its limit within 2 hours or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours. l SURVEILLANCE REQUIREMENTS i 4.2.5.1 Each of the parameters shown above shall be verified to be within its limits at least once per 12 hours. 4.2.5.2 The RCS flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. 1 4.2.5.3 The RCS flow rate shall be demonstrated by measurement once per 18 I months. i , L.
- Limit not applicable during either a THERMAL POWER ramp in excess of 5% of i
RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of , l RATED THERMAL POWER. l l ; L I l- TURKEY POINT - UNITS 3 & 4 3/4 2-16 AMENDMENT N05.137 AND 132 1
l 3/4.3 INSTRUMENTATION ; 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION , i LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.' APPLICABILITY: As shown in Table 3.3-1. ACTION: 1 As shown in Table 3.3-1. SURVEILLANCE REQUIREMENTS _ I 4.3.1.1 Each reactor Trip System instrumentation channel and interlock and i ! the automatic trip logic shall be demonstrated OPERABLE by the performance of 1 the Reactor Trip System Instrumentation Surveillance Requirement specified in I Table 4.3-1. 1 1 t a t i i 1 TURKEY POINT - UNITS 3 & 4 3/4 3-1 AMENDMENT N05.13EAND 132
TABLE 3.3 $
# REACTOR TRIP SYSTEM INSTRUMENTATION A , MINIMUM ; S TOTAL NO. CHANNELS CHANNELS APPLICABLE 5 FUNCTIONAL UNIT OF CHANNELS TO TRIF OPERABLE MODES ACTION c- 1. Manual Reactor Trip 2 1 2 1, 2 l' s'E 2 1- 2 3* , 4 *, 5* 9 d
w 2. Power Range, Neutron Flux
- e. a. High Setpoint 4 2 3 1, 2 2
. b. Low Setpoint 4 2 3 1M, 2 2
- 3. Intermediate Range, Neutron Flux 2 1 2 IN, 2 3
- 4. Source Range, Neutron Flux w a. Startup 2 1 2 2# 4 1 b. Shutdown ** 2 0 2 3,4,5 5 w c. Shutdown 2 1 2 3* , 4*, 5* - 9 0
- 5. Overtemperature AT 3 2 2 1, 2 6
- 6. Overpower AT 3 2 2 1, 2 6
- 7. Pressurizer Pressure-Low 3 2 2 1 6 (Above P-7) ,
i k 8. Pressurizer Pressure--High 3 2 2 1, 2 6 ! a ! 4 9. Pressurizer Water Level--High 3 2 2 1 6 l 5 (Above P-7) - z l 8 10. Reactor Coolant Flow--Low _, a. Single Loop (Above P-8) 3/ loop 2/ loop 2/ loop 1 6 w b. Two Loops (Above P-7 3/ loop 2/ loop , 2/ loop 1 6 l i o and below P-8) U
~ ,
I I
k
.iV .f
,, TABLE 3.3-1 (Continued) C
". REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM i S TOTAL NO. CHANNELS CHANNELS APPLICABLE 5 FUNCTIONAL UNIT. OF CHANNELS TO TRIP . OPERABLE MODES ACTION l- C
- 11. Steam Generator Water 3/sta. gen. 2/sta. gen. 2/sta. gen. 1, 2 6 l,- Level--Low-Low m
w 12. Steam Generator Water Level-- 2 stm. gen. I sim. gen. 1 sim. gen.-1, 2 6 i e- Low Coincident With Steam / 1evel and level coin- level and l
- Feedwater Flow Mismatch 2 sta./ feed- cident with 2 sta./fetM-water flow 1 sta./ feed- water flow mismatch in water flow mismatch in i 'each stm. gen. mismatch in same sim. gen.
same sim. or 2 sta. gen. w gen. 1evel and 1 l 1 sta./feedwater I w flow mismatch
/> in same sta.
l gen.
- 13. Undervoltage--4.16 KV Busses 2/ bus 1/ bus on 2/ bus 1 12 A and 8 (Above P-7) both busssi
- 14. Underfrequency-Trip of Reactor 2/ bus 1 to trip 2/ bus 1 11 i g Coolant Pump Breaker (s) Open RCPs***
g (Above P-7) a 5 15. Turbine Trip (Above P-7) l 5 a. Autostop Oil Pressure 3 2 2 1 12 z b. Turbine Stop Valve Closure 2 2 2 1 12 8 L E o G3 N
, , . . . . . _ _ , . , --_ _ , . - ~ .
TABLE 3.3-1 (Continued) R REACTOR TRIP SYSTEM INSTRUMENTATION Q
, MINIMUM S TOTAL NO. CHANNELS CHANNELS APPLICABLE 5 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ' ACTION h 16. Safety Injection Input 5 from ESF 2 1 2 1, 2 8 d
w 17. Reactor Trip System Interlocks
- e. a. . Intermediate Range
. Neutron Flux, P-6 2 1 2 2# 7
- b. Low Power Reactor Trips Block, P-7 P-10 Input 4 2 3 1 7
. or w Turbine First 2 1 2 1 7 1 Stage Pressure w i c. Power Range Neutron Flux, P-8 4 2 3 ' 1 7
- d. Power Range Neutron Flux, P-10 4 2 3 1, 2 7
- 18. Reactor Coolant, Pump Breaker -
Position Trip 3 a. Above P-8 1/ breaker 1 1/ breaker 'l 11
% B. Above P-7 and below P-8 1/ breaker 2 1/ breaker 1 11 5
M 19. Reactor Trip Breakers 2 1 2 1, 2 8, 10 5 2 1 2 3*, 4*, 5* 9 h 20. Automatic Trip and Interlock 2 1 2 1, 2 8 g Logic 2 1 2 3*, 4*, 5* 9 w 5 _ - *' w v w _ . - _ . _e___._._______-_____________.________.-_..____2_m.__r__ m._ ___ =-.._m _ _ _ m. m ._-_- - .m,___2-__ .mm.__.-
4 TABLE 3.3-1 (Continued) TABLE NOTATION
*When the Reactor Trip System breakers are in the closed position and the Control Rod Drive System is capable of rod withdrawal.
**When the Reactor Trip System breakers are in the open position, one or both of this require-the backup ment. NIS instrumentation For backup channelssee NIS testing requirements, may be used to3satisfy Specification /4.3.3.3, ACCIDENT MONITORING.
*"* Reactor Coolant Pump breaker A is tripped by underfrequency sensor UF-3Al (UF-4A1) or UF-3B1(UF-4B1). Reactor Coolant Pump breakers B and C are tripped by underfrequency sensor UF-3A2(UF-4A2) or UF-382(UF-4B2).
#Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint. 1
##Below the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
1 ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum - Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the I next 6 hours.
- ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
l a. The inoperable channel is placed in the. tripped condition within 1 hour, i
- b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for_up to 2 hours for surveillance testing of other channels per per Specifica-tion 4.3.1.1, and l c. Either, THERMAL POWER is restricted to less than or equal l to 75% of RATED THERHAL POWER.and the Power Range Neutron l Flux Trip Setpoint is reduced to less than or equal to l 85% of RATED THERMAL POWER within 4 hours; or, the QUADRANT l POWER TILT RATIO is i.1onitored per Specification 4.2.4.2.
1 l 132 TURKEY POINT - UNITS 3 & 4 3/4 3-5 AMENDMENT NOS.137AND ,
TABLE 3.3-1 (Continued) ACTION STATEMENTS (Continued) ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:
- a. Below the P-6 (Intermediate Range Neutron Flux Interlock)
Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint, and
- b. Above the P-6 (Intermediate Range Neutron Flux Interlock)
Setpoint but below 10% of RATED THERMAL POWER, restore the inopereie channel to OPERABLE status prior to increasing TH GMAL POWER above 10% of RATED THERMAL POWER. ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving - positive reactivity changes. , ACTION 5 - With the number of OPERABLE channels one less than the Minimum ' Channels OPERABLE requirement, suspend all operations involving positive reactivity changes and verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as appli-cable, within I hour and at least once per 12 hours thereafter. ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required ANALOG CHANNEL OPERATIONAL TEST'provided the inoperable channel is placed in the tripped condition within 1 hour. l ACTION 7 - With less than the Minimum Number of Channels OPERABLE, witnin 1 hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3. ACTION 8 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within ; 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE. ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or open the Reactor Trip System breakers within the next hour. ACTION 10 - With one of the diverse trip features (undervoltage or shunt trip . attachment) inoperable, restore it to OPERABLE status within 48 hours or declare the breaker inoperable and apply ACTION 8. The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing mainten-ance to restore the breaker to OPERABLE states. TURKEY POINT - UNITS 3 & 4 3/4 3-6 AMENDMENT N05.137AND:132
-?-
- TABLE 3.3-1 (Continued)
ACTION STATEMENTS (Continued) ACTION 11 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be.in at least HOT STANDBY within 6 hours. ACTION 12 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required ACTUATION LOGIC TEST provided the inoperable channel is placed in the tripped condition within 1 hour.
)
e t TURKEY POINT - UNITS 3 & 4 3/4 3-7 AMEN 0 MENT N05.137AND 132 - +
TABLE 4.3-1 5
'A
~
REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS T u TRIP S ANALOG ACTUATING MODES FOR
'i CHANNEL DEVICE WHICH
. CHANNEL CHANNEL OPERATIONAL OPERATIONAL . ACTUATION SURVEILLANCE g FUNCTIONAL UNIT CHECK- CALIBRATION TEST TEST- LOGIC TEST IS REQUIRED 3 1. Manual Reactor Trip M.A. N.A. N.A. R(11) N.A. 1, 2, 3* , 4 * , 5*
- e. 2. Power Range, Neutron Flux
= a. High Setpoint S D(2, 4), M N.A. M.A. 1, 2 M(3,4),
Q(4,6), R(4) w b. Low Setpoint S R(4) M N.A. N.A. 1***, 2 1 w 3. Intermediate Range, S. R(4) S/U(1),M N.A. M.A. 1***, 2 s Neutron Flux
- 4. Source Range, Neutron Flux 5 R(4) S/U(1),M(9) fl. A. M.A. 2**, 3, 4, 5
- 5. Overtemperature AT S R(12) M N.A. M.A. 1. 2
- 6. Overpower AT S R M N.A. N.A. 1, 2 k 7. Pressurizer Pressure--Low S R M N.A. N.A. 1 5
4 8. Pressurizer Pressure--High S R M N.A. N.A. 1, 2 z 9. Pressurizer Water Level--High 5 R M N.A. N.A. 1 8 _., 10. Reactcr Coolant Flow--Low 5 R M N.A. N.A. I t'. 3; 11. Steam Generator Water Level-- S R M N.A. M.A. 1, 2 Low-Low C N
TABLE 4.3-1
~
REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS L o TRIP l 8 ANALOG ACTUATING MODES FOR - L "i ~ CHANNEL DEVICE 'WHICH L i Cri1NNEL CHANNEL OPERATIONAL- OPERATIONAL ACTUATION SURVEILLANCE l j FUNCTIONAL UNIT CHECK CALIBRATION ' TEST TEST- LOGIC TEST IS REQUIRED
- l. Z
- 12. Steam Generator Water S R M. N.A. N.A. 1, 2 l
" Level--Low Coincident with l
Steam /Feedwater Flow [ Mismatch
- l. 13. Undervoltage - 4.16 kV N.A. R N.A. N.A.- N.A. 1 Busses A and 8
- 14. Underfrequency - Trip of N.A. R N.A. N.A. N.A. 1 Reactor Coolant Pump
{ Breakers (s) Open w 15. Turbine Trip E a. Autostop Oil Pressure N.A. R N.A. 5/U(1, 10) N.A. 1
- b. Turbine Stop Valve Closure N.A. R N.A. S/U(1, 10) N.A. .1
- 16. Safety Injection Input. ^'
from ESF N.A. -N.A. N.A. .R N.A. 1, 2
- 17. Reactor Trip System
> Interlocks ifi a. Intermediate Range g Neutron Flux, P-6 N.A. R(4) M N.A. N.A. 2**
E
- b. Low Power Reactor Trips Block, F-7 'N.A. R(4) M(8) N.A. N.A. 1-5 (includes P-10 input F and Turbine First y Stage' Pressure)~
3E c. Power Range Neutron F1ux, P-8 N.A. R(4) M(8) N.A. N.A. 1 N$ t
-- - - - - _. ________=_____________:_____________ . .- . - - . . . - - _ - - _ __
_ :u -
- s. -'
- g. .
.~-
- _ y - -
1
- ' TABLE-- 4. 3-1 ~
^
lll - REACTOR-TRIP SYSTEM INSTRUNENTATION SURVEILLANCE REQUIREMENTS
'Q
- o. TRIP S :
ANALOG' ACTUATING MODES FOR 5 CHANNEL? . DEVICE ' WHICH-
, ' CHANNEL- CHANNEL .- OPERATIONAL OPERATIONAL ACTUATION- : SURVEILLANCE:
c FUNCTIONAL UNIT CHECK ' CALIBRATION TEST ' -' TEST- : LOGIC TEST IS REQUIRED ~ j'i . d 17. Reactor Trip System Interlocks (Continued) e d. Power Range
. Neutron Flux, P-10 N.A. .R(4) M(8) N.A. N.A. -1, 2
, 18. Reactor Coolant Pump N.A. N.A. N.A. R N.A. 1 Breaker Pcsition Trip g 19. Reactor Trip Breaker N.A. N.A. M.A. M(7,11) N.A. 1, 2, 3*,-4*, 5* Y 20. Automadc Trip and Inter-O lock Logic N.A. N.A. N.A. N.A. M(7,14) 1, 2, 3*,.4*, 5*~
- 21. Reactor Trip Gypass Breaker N.A. N. A. N. A. M(13),R(15) N.A. "1, 2, 3*, 4*, 5* ,
i V E E . E ' g :. .. If 5 C m
., . , , , , . g- -m --,,,1 .,=e-- =e- .g .e.--- v -e
- a- " =w:*
,. , 3 .
wef 3m - e--- **-&e +P'- e w* ev-=+- e *+e i+t --
'W-V**W'*
- ~
l b c TABLE 4.3-1 (Continued) TABLE NOTATIONS
*When the Reactor Trip System breakers are closed and the Control Rod Dr've System is capable of rod withdrawal.
**Below P-6 (Intermediate Range Neutron Flux-Interlock) Setpoint.
***Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
(1): If not performed in previous 7 days.
. (2). Comparison of calorimetric to excore power indication above 15% of' RATED THERMAL POWER. Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of 4
Specification 4.0.4 are not applicable to entry into MODE 2 or 1.. (3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1. (4) Neutron detectors may be excluded from CHANNEL CALIBRATION. (5) This table Notation number is not used.
'(6)-.Incore-Excore Calibration, above 75% of RATED THERMAL POWER (RTP). If the quarterly surveillance requirement coincides with sustained operation between 30% and 75% of RTP, calibration shall be performed at this lower power level. The provisions of Specification 4.0'.4 are not applicable for entry into MODE 2 or 1.
(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. (8) With power greater than or equal to the Interlock Setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the interlock is in the required state by observing the permissive annunciator window. (9) Monthly surveillance in MODES 3*, 4*, and 5* shall also include verifica-tion that permissive P-6 and P-10 are in their required state for exist-ing plant conditions by observation of the permissive annunciator window, Monthly surveillance shall include verification of the High Flux at Shut-down Alarm Setpoint of 1/2 decade above the existing count rate. (10)'Setpoint verification is not applicable. (11) The TRTP ACIuAiiNO DEVICE OPERATICNAL TEST shall include independent varification of the OPE,% ILITY of the undervoltage and shunt trip attachment of the Reactor Trip Bra akers. TURKEY POINT - UNITS 3 & 4 3/4 3-11 AMENDMENT NOS.137 AND 132 4
.f Ju , .;
g t TABLE 4.3-l'(Continued) gl H o- TABLE NOTATIONS-1
, (12) CHANNEL' CALIBRATION shalljinclude the RTD bypass loops flow rate.
g; (13) Remote manual ~undervoltage trip when breaker placed in service.
; {14): Interlock 1.ogic Test shall consist of verifying that the: interlock is in
.its reopired state by observing the permissive annunciator window.
(15) AutorAtic undervoltage trip. A e 1 (i I.~ 3 e I
'[
, 4 -
y,,'
,i
\
f ,N ,
.- \ 9
~l; .
~ ' 01 N.
[.!t 6 lv ; ti >
\i' <1
, iTURKEY POINT 'JNITS 3 & 4 3/4 3-12 AMENDMENT NOS.137 AND 132 a a: s
a INSTRUMENTATION ,
. 3/4.3.2 ENGINEERED SAFETY FEATURES' ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION r-
~
- 3.3.2. The Engineered Safety Feature Actuation System (ESFAS) instrumentation '
, channels and interlocks shown in Table 3.3-2 shall be OPERABLE with their Trip
. Setpointsiset consistent with the values shown in the Trip Setpoint column of iTable 3.3-3..
- APPLICABILITY: As shown in Table 3.3-2.
ACTION:- a.. 1With an ESFAS-Instrumentation or' Interlock Trip Setpoint trip less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.5 3, adjust the Setpoint consistent with the Trip Setroint < value.
- b. With an ESFAS Instrumentation or Interlock Trip Setpoint.less conservative than the value shown in the Allowable.Value column of Table 3.3-3, declare the channel inoperable and apply-the applicable ACTION statement requirements of Table 3.3-2 until the' channel.is restored to OPERABLE status with its Setpoint adjusted consistent-with.the Trip Setpoint value.
c._ With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-2. SURVEILLANCE REQUIREMENTS 4.3.2.1 Cach ESFAS instrumentation channel and interlock and the automatic : actuat bn logic and relays shall be demonstrated OPERABLE by performance of the E5FAS Instrumentation Surveillance Requirements specified in Table 4.3-2. 1 1 0 4 TURKEY POINT - UNITS 3 & 4 3/4 3-13 AMENDMENT N05.137AND 132' a
e-- g, r . _., - TABLE 3.3-2
- c j - ENGINEERED SAFETY FEATURES ~ ACTUATION SYSTEM INSTRUMENTATION Q
, +
- l. S MINIMUM i 5 TOTAL'N0.
CHANNELS CHANNELS APPLICABLE
- , FUNCTIONAL UNIT OF CHANNELS- TO TRIP OPERABLE MDDES ACTION-
< c - .
. . ~ "
i'i
- 1. Safety Injection (Reactor -
d Trip, Turbine Trip, Feedwater w Isolation, Control Room
- o. - Ventilation Isolation, Start 4 . Diesel. Generators, Contain-ment Phase A. Isolation (ex- '
cept Manual SI), Containment Cooling Fans, Containment. Filter Fans, Start Sequencer, R Component Cooling Water, Start z'
- Auxiliary Feedwater and Intake y Cooling Water). --
- a. Manual Initiation 2 1 2 1, 2, 3, 4 17
- b. Automatic Actuation 2 1 2 1,2,3,4 14- .,
Logic and Actuation ~ Relays
- c. -Containment 3 2 2 1, 2,~3' 15 k Pressure-High 5
j5; d. Pressurizer 3- -2 2 1, 2, 3# .15-5 -Pressure -' Low z 8 e. High Differential 3/ steam line 2/ steam line 2/ steam 1, 2, 3* 15 g Pressure'Between in any steam .line w the Steam Line
~
^1ine g Header and.any o,
Steam Line M
. . ,, o a ..s~r 4 w
--.o. a - - .- -- -% "-@ m,- -
tr w e' w --ef~r+=- e + - - ' - - - - - - '--u' "- m'----C'--1-=
- - g ,
~
.-TABLE 3.'3-2 (Continued)' ~
y ~ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.
- h. '
Q ~ MINIMUM-CHANNELS : CHANNELS APPLICABLE 1 TOTAL NO. S . CT CHANNELS -TO TRIP OPERARLE MODES - ' ACTION
$ FUNCTIONAL-UNIT
~
1/ steam line 1/ steam line. 1, 2, 3* 15 Steam Line Flow--High' -2/ steam line h z f. Coincident with; in any two in any two steam lines steam lines g w
, Steam Generator .1, 2,-3* 15 Pressure--Low 1/ steam 1/ steam line 1/ steam-
. generator in any two generator steam lines in any two
- steam lines
~
or T --Low 1/ loop 1/ loop in any 1/ loop in any .1, 2, 3* 15 -- R yg two:1 oops two loops w
- 2. Containment Spray h
1 -2 1, 2, 3, 4 14
- a. Automatic Actuation 2 Logic and Actuation Relays 3 2 2 1, 2,.3 15. '-
- b. Containment Pressure--
High-High E Coincident with: I h Containeent Pressure-- 1,2,3 15-3 2 '2 g High
'i 3. Containment Isolation z
O E a. Phase "A" Isolation 2 1,2,3,4 17 ti 1) Manual Initiation '2 1~ 2 - 1,2,3,4 14 Automatic Actuation 2 1 g 2) Logic and Actuation
- o -
d Relays to 4-S
' cM , , , , , , , . . +
Q~'
- TABLE 3.3-2 (Continued)-
g ENGINEERED SAFETY FEATURES: ACTUATION SYSTEM INSTRIMENTATION Q MINIMUM o TOTAL NO. CHANNELS CHANNELS ' APPLICABLE S 0F. CHANNELS TO-TRIP OPERABLE _ MODES LACTION. _ _ j
$ FUNCTIONAL UNIT J h 3. Containment Isolation (Continued) a Safety Injection See Item 1. above for all Safety Injection initiating fuxtions and '
i;! 3) requirements. -(Manual S.I. initiation will'not' initiate Phase A w e.
' Isolation.)
.sb
- b. Phase "B" Isolation 1,2y3,4 .17-Manual Initiation 2 2 (Both 2
'1) buttons must ._
be pushed simultaneously R
# to' actuate) w 14 2 1 2 1,2,3,.4~
h 2) Automatic Actuation Logic and Actuation Relays 1, 2, 3 . 2 2 15
- 3) Containment ~3 Pressure--High-High Coincident with:
g Containment Pressure-- ~ 1 , 2 ,- 3 15 3 2 2 g High x ~ 3 c. Containment Ventilation Isolation o E 1) Containment Isola- See Items 3.a.1 and 3.b.1'above for all Manual Containment
' tion Manual Phase A Ventilation functions and requirements.
U p or Manual Phase B' S _. . N . e4
~ ~
~
=
ls y TABLE 3.3-2 (Continued) - x h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTA"] 0
, MINIMUM-
- o. TOTAL NO. CHANNELS CHANNELS- APPLICABLE
$ ' FUNCTIONAL UNIT OF CHANNELS TO' TRIP - OPERABLE MODES ACTION-
- 3. Containment Isolation (Continued) 5 l d. 2) Automatic Actuation 2 .1 2 -1,2,3.4 16-w Logic and Actuation Relays
- 3) Safety Injection See Item 1.-above for all Safety Injection initiating functions requirements.
- 4) Containment 2## 'l 1. 1,2,3,4 16 R Radioactivity-High _
w 4. Steam Line Isolation
- a. Manual Initiation 1/ operating 1/ operating 1/ operating 1,2,3 21 (individual) steam line steam line steam line
- b. Automatic 2 1 2 1,2,3 20 Actuation Logic and Actuation-Relays
$ c. Containment Pressure-- 3 2 2 1,2,3 15 at High-High E Coincident with:
[ Containment Pressure-- ~3 2 -2 1,2,3 15 g High C w 5 0
~
4 a4cm . -.q. _.%-+_ v. f. A ,, w = . . _ _ _ _ _ _ _ _ _ _ _ _ _ , . _ _ _ , _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ -
..i.e.g gg , 4 , m . ..
'- + w g. e ,a
, rg
- y. . TABLE 3.3-2 (Continued)
- c. _
.R ~ ENGINEERED St.FETY-FEATURES ACTUATION SYSTEM INSTRUMENTATION' -
Q ~
, MINIMUM S --
' TOTAL N0. CHANNELS CHANNELS. APPLICABLE
~
~
FUNCTIONAL UNIT OF-CHANNELS- TO TRIP - OPERABLE M00ES ACTION-
- 4. Steam Line~ Isolation (Continued) 5 .
.s d d. Steam Line Flow--High' 2/ steam line-- 1/ steam'line 1/ steam line 1, 2, . 3 15 -
w Coincident with: 3 Steam Generator
, ' Pressure--Low 1/ steam 1/ steam 1/ steam 1, 2, 3 15 generator ' generator . generator in any two in any two steam lines' steam lines or R' T,y9--Low 1/ loop 1/ loop in - 1/ loop in ~ 1, 2, 3 '15 --
- any two any two
~ Y - 1 oops . 1 oops 5
- 5. Feedwater Isolation
- a. Automatic Actua- 2 1 2 :- 1, 2 . 22 tion Logic and -*
Actuation Relays
~ > b. Safety-Injection See Item 1. above for all Safety Injection initiating functions 3 and requirements.
5 g 6. Auxiliary Feedwater### 2 a. Automatic Actua- 2 1 2 1,2,3 - 20 g tion Logic and
- Actuation Relays ,
tl 2 c, w
-~"---._n____=_.___a-
- x. _.
_..--_____._._n.____.-_ > __---___ r _--_-x--.x._____.-- - _ - -*_ . - -- mn:u_ m., , a. _ n <s- - ..a '
.m.. - .a . - - -
~
TABLE 3.3-2 (Continued) _ ENGINEERED' SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION fG MINIMUM -
, -CHANNELS CHANNELS APPLICABLE TOTAL NO.
S OF CHANNELS' TO TRIP' OPERABLE M00ES -ACTION , 2j FUNCTIONAL UNIT
- 6. Auxiliary Feedwater### (Continued) 2/ steam 1, 2, 3 15
- b. Sta. Gen. Water Level-- 3/ steam 2/ steam 3 Low-Low generator generator generator
. m in any-e steam a generator
- c. Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.
'1/ bus 1,2,3- 23
- d. Bus Stripping 1/ bus 1/ bus w
0
- e. Trip of-All Main Feed- _
(1/ breaker) (1/ breaker) 1, 2 23' water Pumps Breakers 1/ breaker -
/ operating / operating pump pump
- 7. Loss of Power 2/ bus 1, 2, 3, 4 18
- a. 4.16 kV' Busses A and B 2/ bus 2/ bus 5 (Loss of Voltage)
E 2 on'any 2 per load 1, 2, 3, 4- 18' E b. 480 V Load Centers 2 per load 3A, 3B,'3C, 3D and center load center center E
" 4A, 4B, 4C, 4D 5 (2 instantaneous P relays per load d center) g o
Degraded Voltage Coincident with: See Item 1. above'for all~ Safety Injection' initiating functions a Safety Injection and-requirements M IIll l
.; r - 3 -<
~ '
q . 1: ,
~
~
y TABLE 3.3-2-(Continued)-
- k. .. ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION-Q. '
y , MINIMUM - S ETOTAL NO. _ CHANNELS' CHANNELS APPLICABLE mi FUNCTIONAL UNIT OF CHANNELS TO' TRIP. OPERABLE- ' MODES ACTION _ c- 7. Loss of Power (Continued) d c. 480 V Load Centers 2 per load 2 on any 2 per load 1, 2, 3,14, 18 w - 3A, 3B, 3C, 3D and center. load center' center
- g. 4A, 4B, 4C 4D '
, (2 inverse time relays per load.
center) Degraded Voltage w 8. Engineered Safety Features 1 Actuation System Interlocks w a. Pressurizer Pressure 1,2,3 3 2 2 19. El
- b. T,yg - Low 3 2 2 -1, 2, 3 19- ,
- 9. Control Room Ventilation Isolation. -
- a. Automatic Actuation 2- .1 2 1, 2, 3, 4,6** 16 Logic and Actuation ~ '
3 Relays h b. Safety Injection See Item 1. above for all Safety Injection' initiating functions c y and requirements.
$ c. Containment Radio- 2 1 1 1, 2, 3,4,6**-16 2 activity--High .O d. Containment Isolation 2 1 2 1,2,3,4- 17-O Manual Phase-A or y Manual.-Phase B~
5 e. Control Room Air 2 'l 2 'All 24 C Intake Radiation-
" Level S
~1..___- __ c_.-_ ~ ~ - , ax . - - - .a .e .ew. + , - -
s i 1 TABLE-3.3-2 (Continued)" TABLE NOTATION
# Trip'. function.may'be blocked in this MODE below the Pressurizer Pressure Interlock Setpoint of 2000 psig. =
## Channels are for particulate' radioactivity.and for gaseous radioactivity. '
'### Auxiliary feedwater manual initiation is included in Specification '3.7.1.2. '
'[
- Trip function may be blocked in'this MODE below the T --Low Interlock Setpoint. avg
~
**0nly during CORE ALTERATIONS or movement of irradiated fuel within the ,
containment,
, ACTION' STATEMENTS. . ACTION 14 - than the Minimum With the number Channels OPERABLE of OPERABLE channels requirement, be in at leastone less, HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1, provided j the other channel is OPERABLE.
ACTION 15 - Withthenumber'ofOPERABLEchanneIsonelessthantheTotal Number of Channels, operation may proceed until' performance of the next required ANALOG CHANNEL OPERATIONAL TEST or TRIP
' ACTUATING DEVICE OPERATIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour.
ACTION 16 - With less than the Minimum Channels 0PERABLE requirement, comply with the ACTION statement requirements of Specifica-
- tion 3.3.3:1 Item la of Table 3.3-4. ,
ACTION 17 - With the number of OPERABLE channels one less than the Minimum 1 Channels OPERABLE requirement, restore the inoperable channel l to OPERABLE status within 48 hours or be in at least HOT STANDBY < within the next 6 hours and in~ COLD SHUTDOWN within the following 30 hours. 4
?
I TURKEY POINT - UNITS 3 & 4 3/4 3-21 AMENDMENT N05.137AND 132 l
w i 9 TABLE-3.'3-2 (Continued) TABLE NOTATION (Continued) ,
- ACTION 18 - With t he number of OPERABLE channels one. less than the Total' Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the inoperable channel is placed in the tripped h condition within 1 hour.
L
- ACTION 19 - With less than the Minimum Number of Channels OPERABLE, within I hour determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant. condition, or apply Specifica-t tion 3.0.3.
L ACTION 20 - With the number of OPERABLE channels one'1ess than the Minimum L Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6. hours; L however, one channel may be bypassed for up to 2 hours for ;
- l. surveillance testing per Specification 4.3.2.1 provided the. t E other channel is OPERABLE.
F - ACTION 21 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE-status within 48 hours or declare the associated valve inoper-p able and take the ACTION required by Specification 3.7.1.5. 1 L ACTION 22 - With the number of OPERABLE channels one less than the Minimum-
- i. Channels OPERABLE requirement, be in at least HOT STANDBY l- within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.2.1 provided the other channel is OPERABLE.
ACTION 23 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, comply with Specification 3.0.3. . -ACTION 24 - With'the number of OPERABLE channels one less than the Minimum L Channels OPERABLE requirement, within 1 hour isolate the control i room Emergency' Ventilation System and initiate operation of the - l Control Room Emergency Ventilation System in the recirculation .. mode. L TURKEY POINT - UNITS 3 & 4 3/4 3-22 AMENDMENT N05.137AND 132
y s ;
' i
__,p TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS j
-i TRIP FUNCTIONAL UNIT SETPOINT ALLOWABLE VALUE# r
- 1. . Safety Injection (Reactor Trip, Turbine Trip, Feedwater Isolation, ,
Control Room Ventilation Isolation, ' Start Diesel: Generators, Contain-t- ment Phase A Isolation (except . Manual SI), containment Cooling Fans, Containment Filter Fans, Start Sequencer, Component Cooling Water, Start Auxiliary Feedwater and Intake Cooling Water).
- a. Manual Initiatton N.A. N.A. J b.. Automatic Actuation Logic N.A. N.A.
- c. Containment Pressure--High 16 psig 1[ ] psig
- d. Pressurizer Pressure--Low 11715 psig 1[ ] psig
- e. High Differential Pressure 1150 psi $[ ] psi 1 Between the Steam Line J
. Header and any Steam Line.
- f. Steam Line Flow--High SA . function defined [ ]
as follows: A op .; corresponding to 'l 0.64 x 10s lbs/hr I at 0% load increa-ing linearly to a op corresponding to 3.84~x 108 lbs/hr at full load. Coincident with: 1600 psig 1[ -) psig Steam Generator Pressure--Low or
.T gg --Low 1531*F 1[ ]'F
- 2. Containment Spray .
- a. Automatic Actuation Logic N.A. N.A.
and Actuation Relays
- b. Containment Pressure--High- 130.0 psig 1[ ] psig
- High Coincident with: Containment Pressure--High 16.0 psig $[ ] psig TURKEY POINT - UNITS 3 & 4 3/4 3-23 AMENDMENT NOS.137AND 132 _. _ . _ _ . _ . _ _ _ . _ _ , . #f - e
- j TABLE 3.3-3 (Continued):
s 1 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
- INSTRUMENTATION TRIP SETPOINTS q
TRIP FUNCTIONAL UNIT SETPOINT ALLOWABLE VALU_E# ; [
"1 3.: Containment Isolation
- a. Phase "A" Isolation l t 1)- Manual Initiation N.A. N.A. , ;
- 2) Automatic Actuation Logic N.A. N.A.
1 and-Actuation Relays L 3) Safety Injection See Item 1~above for all Safety 2 " Injection Trip Setpoints and Allowable Values. b.- Phase "B" Isolation
- 1) Manual ~ Initiation N.A. N.A.
- 2) Automatic Actuation Logic N.A. N.A.
and. Actuation Relays l 3) Containment Pressure-- 130.0 psig $[ ] psig L High-High i Coincident with- ! Containment Pressure--High $6.0 psig 1[ ] psig 1
- c. Containment Ventilation Isolation )
L 1) Containment Isolation' N.A. N.A. Hanual Phase A or Manual L E Phase B j .. H 2) -Automatic Actuation Logic N.A. N. A. lf i and Actuation Relays J I
- 3) Safety Injection See Item 1. above for all Safety i Injection' Trip Setpoints and 4 Allowable Values. l l
L 4) Containment Radio- Particulate (R-11) [ ] I L activity--High (1) <6.1 x 105 CPM , l Easeous (R-12) l I See (2) l l l. l TURKEY POINT - UNITS 3 & 4 3/4 3-24 AMENDMENT NOS.?37AND132 'l l l
k',
?+
t , TABLE 3.3-3 (Continued) [ 3 - 1 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
-INSTRUMENTATION TRIP SETPOINTS-1 1
TRIP FUNCTIONAL UNIT- SETPOINT ALLOWABLE VALUEL r ' 4. Steam Line Isolation
- a. Manual Initiation N.A. N.A
- b. Automatic Actuation Logic S. A. N. A'.
and Actuation Relays ,
)
- c. Containment Pressure--High- 130.0 psig 1[ ] psig i High Coincident with:
Containment Pressure--High 16.0 psig $[ ] psig
- f. Steam Line Flow--High 1A function defined [ ].
o as follows: A Ap corresponding to O.64 x 108 lbs/hr-at 0% load increa-ing linearly to a -l Ap corresponding to = 3.84~x 108 lbs/hr. ' at full load. l Coincident with: 1600 psig g[ ]'psig
-Steam Line
- . Pressure--Low ;
or T,yg--Low 1531'F 1[ ]'F
- 5. -Feedwater Isolation:
i
- a. Automatic Actuation' Logic N.A. N.A.
and Actuation Relays ,
- b. Safety Injection See Item 1. above for all Safety .
Injection Trip Setpoints and Allowable Values.
- 6. Auxiliary Feedwater (3)
- a. . Automatic Actuation Logic N.A. N.A.
.and Actuation Relays- i
- b. Steam Generator Water 115% of narrow 1[ ]% of narrow L Level--Low-Low ringe instrument range instrument L span, span.
- c. Safety Injection Se Item 1. above for all Safety l
Injection Trip Setpoints and 1-Allowable Values. l TURKEY POINT - UNITS 3 & 4 3/4 3-25 AMENDMENT NOS.137 AND 132
, c TABLE 3.3-3 (Continued)-
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS
- TRIP
, FUNCTIONAL UNIT SETPOINT ALLOWABLE VALUE#
G. Auxiliary Feedwater (Continued) e
- d. Bus Stripping See Item'7. below for all Bus Stripping Setpoints and Allowable Values,
- e. Trip of All Main Feedwater N. A. = N.A.
Pump Breakers. 7 .' Loss of Power
- a. 4.16-kV Busses A and B N.A. N.A.
(Loss _of Ybitage)
- b. 480V' Load' Centers
-(Instantaneous Relays)
Degraded. Voltage Load Center
]
~
3A 418ViSV.(10 sec i 1 sec delay) -[ 3B 423ViSV (10 see i 1 sec delay) [ ] 3C 429Vt5V (10 sec
- 1 sec delay) [ ]
3D 429VtSV (10 sec i I sec delay) [ ] 4A 407ViSV_(10 sec i i sec delay) [ ] 4B 423ViSV (10 sec i 1 sec delay) [ ] 4C 419VtSV (10 sec i 1 sec delay) [ ] 4D 404V15V (10 sec
- 1 sec delay) [ ]
Coincident with: Safety Injection and See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values. Diesel Generator N.A. N.A. Breaker Open TURKEY POINT - UNITS 3 & 4 3/4 3-26 AMENDMENT NOS.137 AND 132
1
, hj: 4 i
TABLE 3.3-3 (Continued) lQiGINEEREDSAFETYFEATURESACTUATIONSYSTEM INSTRUMENTATION TRIP SETPOINTS TRIP : FUNCTIONAL UNIT SETPOINT ALLOWABLE VALUE#
- 7. Loss of Power (Continued)
- c. 480V Load Centers
, (Inverse Time Relays) t
, Degraded Voltage. ,
Load Center
, c 3A 41CVASV(60 sec 130 [' ]
-sec delay)l
. 38 426VtSV(60 sec f30 [ ]
sec' delay) 3C 436Vi5V(60 see 130 [ ] l sec delay) 3D 437ViSV(60 see 130 [ ]' sec delay) 4A 424Vf5V(60 see i30 [ ] "- sec delay) 48 422ViSV(60 sec 130 [ ] sec delay)-
- ' 4C 433V15V(60'sec 130 [ ]
sec delay) 40 432VtSV(60 sec't30 [ ] l
-sec delay)
Coincident with: " , Diesel Generator Breaker Open N.A. N. A. i
- 8. Engineering Safety Features Actuation System Interlocks-L
- a. Pressurizer Pressure 12000 psig <[
] psig .
- b. .T,yg--Low >531*F
[ ]
- 9. Control Room Ventilation Isolation .
- a. Automatic Actuation Logic and Actuation Relays N.A. N.A. )
1
- b. Safety' Injection See Item 1. above for all Safety-L Injection Trip Setpoints and Allowable Values.
- TURKEY POINT - UNITS 3 & 4 3/4 3-27 AMENDMENT N05q137 AND 132
- n ;
TABLE 3.3-3 (Continued) - ENGINEERED SAFETY FEATURES' ACTUATION-5YSTEM INSTRUMENTATION-TRIP SETPOINTS TRIP FUNCTIONAL UNIT- SETPOINT ALLOWABLE VALUE#.
- 9.= Control Room Isolation (Continued) ,
- c. Containment Radioactivity-- Particulate (R-11) [ -) 9 High (1) <6.1 x 105 CPM -
3aseous (R-12) See-(2)
- d. Containment Isolation. N.A. N. A.
, Manual Phase A or Manual .
Phase B
- e. Air' Intake Radiation Level < 2 mR/hr 2.83 mR/hr TABLE NOTATIONS (1) Either the particulate or gaseous channel in the OPERABLE status will satisfy this LCO.
. (2) Containment. Gaseous Monitor Setpoint = (3.2 x 104) CPM, -
( F ) l Where F = Actual Purae Flow
' Design Purge Flow (35,000 CFM)
Setpoint may vary according to current plant conditions provided that the' release rate does not exceed allowable limits provided in Specification 3.11.2.1. -t (3) Auxiliary feedwater manual initiation is included in Specification 3.7.1.2.
#If no. allowable value is specified so indicated by [ '), the trip setpoint shall also be the allowable value.
l TURKEY POINT - UNITS 3 & 4 3/4 3-28 AMENDMENT N05.137 AND 132 i 1 y a . . _ . _ , . , . ,
a - ~ ~
' TABLE 4.3-2 y ... ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ~
~ g. - SURVEILLANCE PEQUIREMENTS N
-< TRIP g
~
ANALOG ACTUATING MODES CHANNEL DEVICE FOR WHICH
-E CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION. SURVEILLANCE i
FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST' LOGIC TEST # IS REQUIRED-
~
E. 1. Safety Injection (Reactor
.g Trip, Turbine Trip, Feed-
~
, water Isolation, Control ..
, Room Ventilation Isolation,
, Start Diesel Generators.
- Containment Phase A Isola-tion.(except Manual SI),
Containment Cooling Fans, Containment Filter Fans, y Start Sequencer, Component a Cooling Water, Start Auxiliary w Feedwater and Intake Cooling Water)
~
U$ a. Manual' Initiation N.A. N.A. N.A. R N.A. 1, 2, 3
- b. Automatic Actuation N.A. N.A. N.A. N.A. M(1)- 1, 2, 3(3)
Logic and Actuation Relays
- c. Containment Pressure-- N.A. R N. A'. N.A. M(1) - 1,'2,'3-
, High h d. Pressurizer Pressure-- S R M(5) N.A. N.A. - 1, 2, 3(3) g Low m
$ e. High Differential S R M(5) N.A. N.A. ' 1, 2,'3(3);
z Pressure Between the 8 Steam Line Header and h~ any Steam Line SE f. Steam Line Flow--High ..S R M(5) N.A. : N.A. . 1, 2, 3(3) Coincident with: U Steam Generator , Pressure--Low- S R :M(5) N.A. . M.A. 1,-2,'3(3) or T --Low S .R M(5)' N.A. ~ N.A. .1,:2,3(3) avg ., 't
.i~.,, .
. . . . - _ , ,____.,7___,_.f . _ _ _ __
-p , , _.
,y,_ _,,,_,.,%,,,,,
~-
n ; t.:.. , y ' TABLE 4.3-2 (Continued) e _ g ENGINEERED SAFETY-FEATURES ~ ACTUATION SYSTEM INSTRUMENTATION
-< SURVEILLANCE REQUIREMENTS n
3 TRIP
'E ANALOG ACTUATING MODES ~
CHANNEL -DEVICE FOR E1ICH c CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE-3 m FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST. LOGIC TEST # IS REQUIRED _ m-w 2. Containment Spray A a. Automatic Actuation N.A. N. A. N.A. N.A. M(1) 1,'2, 3,'4 Logic and Actuation _ .g Relays
, b. Containment Pressure-- N.A. R N.A. R M(1) 1, 2, 3 '
s High-High l- Coincident with: { o Containment Pressure-- High N.A. R N.A. R M(1) 1, 2, 3 ,
- 3. Containment Isolation
- a. Phase '.'A" Isolation
- 1) Manual Initiation N.4. N.A. N.A. R N.A. 1,2,3,4 E 2) Automatic Actua- N.A. N.A. N.A. N.A. M(1) 1,2,3,4 5 tion Logic and y /,ctuation Relays a
g 3) Safety Injection See Item 1. above for all. Safety Injection Surveillance Requirements. a b. Phase "B" Isolation w g 1) Manual Initiation ~ N.A. N.A. N.A. R N.A. 1, 2, 3,'4 g; 2) Automatic,Actua- N.A. N . A . N.A. N.A. .M(1) 1, 2, 3, 4
" tion Logic and~
Actuation Relays
.s
+w- --. '*y *-m1 s. C .newv. ,,.w.
a wi
~
~
~
. .TABLE 4.3-2 (Continued) m ~
'E -
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g SURVEILLANCE REQUIREMENTS o TRIP-ANALOG ACTUATING N00ES CHANNEL DEVICE FOR WHICH : CHANNEL OPERATIONAL OPERATIONAL 1 ACTUATION : SURVEILLANCE-g CHANNEL ' CHANNEL CHECK CALIBRATION ; TEST TEST LOGIC TEST # IS REQUIRED'
- q FUNCTIONAL UNIT d 3. Containment Isolation (Continued)
- a. 1, 2, 3 N.A. R. N.A. R' 'M(1)
*o 3) Containment Pressure--High-High Coincident with: Containment 1, 2, 3 -
Pressure--High N.A. R N.A. R M(1)
,, ~
3:
- c. Containment Venti-
[ 1ation Isolation-
- N.A. 1., 2,'3, 4 N.A. N.A. N.A. R
- 1) Containment Isolation Manual' Phase A or. Manual Phase B N.A. N.A. N.A. N.A. N.A.
- 2) Automatic Actua-3; tion Logic and m Actuation Relays o
- 3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.
a R .M N.A. N.A. 1,.2, 3, 4 g 4) Containment Radio- S
~ ~
v> activity--High I3
;[ 4. Steam LineJIsolation-z :R -N.A. 1, 2,'3
N.A. N.A. N.A.
- a. Manual Initiation EI . .
N.A. N.A. M.A. M(1) -1, 2, 3(3)
** b. ~ Automatic Actuation- N.A.
Logic and Actuation Relays
- l. , -:
~ < , , , , , , . . , . , _ _
_. :s u. u _
- :== , . .c
~
- TABLE 4.-3-2'(Continued)' ^
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION h
-< SURVEILLANCE REQUIREMENTS -
TRIP. _ ANALOG ' ACTUATING- MODES-- M 1 DEVICE -FOR WHICH' 8 CHANNEL ~ CHANNEL OPERATIONAL- OPERATIONAL'. ACTUATION ^ SURVEILLANCE g CHANNEL CHANNEL CHECK CALIBRATION TEST TEST LOGIC TEST # IS REQUIRED y FUNCTIONAL UNIT-w 4. Steam Line Isolation (Continued) 9 1, 2, 3 . Containment Pressure-- N.A. R N.A. R M(1) A c. High-High Coincident with: M(1) 1, 2,~3 Containment Pressure-- N.A. R N.A. R-High Steam Line Flow--High M(5) N.A. N.A. 1, 2, 3
- d. S(3) R.
Y Coincident with: U Steam Generator N.A. N.A. 1, 2, 3 Pressure--Low S(3) R M(5) o f' M(5) N.A. M.A. 1, 2, 3 . T,yg--Low S(3) .R _
- 5. Feedwater Isolation N.A. N.A. R 1, 2
> a. Automatic Actuation N.A. N.A.
M Logic and Actuation ~ y Relays m M b. Safety Injection. See Item 1. above for all' Safety Injection. Surveillance Requirements. 2 8, 6. Auxiliary Feedwater (2) N.A. N.A. N.A. R ,1, 2, 3-U a. Automatic Actuation N.A. E Logic and Actuation Relays S R M N.A. N.A. 1,'2, 3
- b. Steam Generator Water Level--Low-Low' 9
~ ' -< - - - - - -
- mI , ,
,,-. ._ w.
y 2 - 3 . x- :m - s
-_ L , -
- w. - -
m . . . g-f,[ [. I ~( y_ . TABLE 4.3-2 (Continued)~ _ g . _ - g- ' ENGINEERED SAFETY FEATURES ACTUATION-SYSTEM INSTRUMENTATION
-< SURVEILLANCE REQUIREMENTS -
o n - ~ g TRIP- .
-4 ANALOG ' ACTUATING: MODES..
a
~ CHANNEL DEVICE ~ FOR WHICH c CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL. ACTUATION SURVEILLANCE-h m ..
FUNCTIONAL UNIT CHECK CALIBRATION - iTEST TEST LOGIC TEST # IS REQUIRED Auxiliary Feedwater (Continued)
~
I w 6. '
- o. ~
A c. Safety Injection See-Item 1. above for all Safety Injection Surveillance Requirements.
- d. Bus Stripping 'N.A. R N.A. R .N.A. 1,2,3.
r_.
, e. Trip of All Main N.A. N.A. N.A. R N.A. 1, 2:
g Feedwater Pump ' '
, Breakers. ,
U 7. Loss of Power <
- a. 4.16 kV Busses A N. A.- R- N.A. R N.A. 1 , .2 , 3 ,5 4 and B (Loss of Voltage) ,
- b. 480V Load Centers S R N.A. M(1). N.A. 1, 2, 3, 4
> 3A,38,3C,3D and 5 4A,48,4C,4D E (Instantaneous 5 Relays) Degraded
'i Voltage '
z 8 Coincident with: i --. ' Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.. I O
'E c. 480V Load Centers- S R ~ N.A. .. M(1) - N.A. 1,-2,:3,L4-
! 3A,3B,3C,3D and
~
C 4A,48,4C,4D -- (Inverse Time . Relays) Degraded .. Voltage -
%- . - - * ~ . . . . e3; s . , _ _ _ _
..__ .._r-4__ <__m--_ _ _ _ _ _ ' * -
- . .- _ _ _ _ _ _ _ . , _ - - - - . - - ..,t2_m _m__-_.
+-.w-we .-+ .t-.e- .m-u-.-. %
w - - E x: -- 2 M 1-ZT ~L [
~
2 "'-
^^vtw '
7^^: m ~ I y ~ TABLE'4.3-2-(Continued)- E -ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION h -SURVEILLANCE REQUIREMENTS- - o TRIP i 8 - ANALOG- ACTUATING MODES 4 . CHANNEL DEVICE ._ FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE g FUNCTIONAL UNIT CHECK CALIBRATION ~ TEST _ TEST -LOGIC TEST # IS REQUIRED 3 8. Engineering Safety , Features Actuation i- [ System Interlocks
= a. Pressurizer Pressure N.A. R M(5) N.A. N.A. 1, 2,-3(3)
- b. T yg --Low N.A. R .M(5) N.A. N.A. 1, 2, 3(3)
- 9. Control Roon Ventilation Isolation -
- a. Automatic Actuation N.A. N.A. N.A. N.A. M.A.
Logic and Actuation
~
y ' g Relays
- b. Safety Inj' action See-Item 1. above for all' Safety Injection Surveillance Requirements..
- c. Containment- S' R M N.A. N.A. (4)
Radioactivity--High
- d. Containment Isolation N.A. N.A. N.A. R N. A. - 1,2,3,4.
Manual' Phase A or g Manual Phase B + E e. Control Room Air 9 R H N.A. N.A. All E Intake Radiation Level
$ TABLE NOTATIONS 5 (1) Each train shall be tested at least every 62 days on a STAGGERED: TEST BASIS.
E (2) Auxiliary feedwater manual initiation is included in Specification 3.7.1.2.
$ (3) The provisions- of Specification -4.0.4 are not applicable for entering Mods.3, provided that the w applicable surveillances are completed within % hours from entering Mode 3. ~ *
(4) Applicable in MODES 1, 2, 3, 4 or during CORE ALTERATIONS or movement of irradiated fuel within the m containment. M (5) Test of alarm function not required when alare locked in.
#At least once per 18 months each Actuation. Logic Test shall include energization of each.
relay and verification.of OPERABILITY of each relay.
+ = - - * ~ _ -- -@_M--.a
.m = &;-+
, , , , ,_ _ __ ,__ _ __ __4 _
_ ~ , , , _ _ _
INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION
-RADIATION MONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels-for plant operations shown in Table 3.3-4 shall be OPERABLE with their Alarm / Trip Setpoints within the specified limits.
APPLICABILITY: As shown in Table 3.3-4. S ACTION:
- a. With a radiation monitoring channel Alarm / Trip Setpoint for plant operations exceeding the value shown in Table 3.3-4, adjust the Setpoint-to within the limit within 4 hours or declare the channel inoperable.
b .- With one or more radiation monitoring channels for plant operations inoperable, take the ACTION shown in Table 3.3-4.
- c. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST for the MODES and at the frequencies shown in Table 4.3-3. TURKEY POINT - UNITS 3 & 4 3/4 3-35 AMENDMENT NOS.137AND 132 P
x. L - p -
- t. -
.c w
~~
TABLE 3.3-4 ~ l -4 i RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS 7 . CHANNELS. CHANNELS ~ APPLICABLE. ALARM / TRIP' S FUNCTIONAL UNIT :TO TRIP / ALARM .: OPERABLE M00ES: -
..SETPOINT ACTION .
~
.. 1. Containment. '
C . 5 a. Containment-Atmosphere 1' 1* - All*- .. Particulate 26 for M00ES-1,:2, 3; 41 i l' d Radioactivity-High <6.1x105 CPM
~
'or~ '
w (Particulate or laseous. ^
- e. Gaseous (See Note L)) See Note 2. 27 M00ES 5 AND 6-
- b. RCS Leakage Detection N.A. .1 1, 2, 3, 4. N.A. ,
26-Particulate Radio-activity or Gaseous Radioactivity k' >
- 2. Spent Fuel Storage Pool Areas -
y T ~
- a. ** <5.5x10 2 pCi M Unit 3 Radioactivity - 1 1 28 High Gaseous 'cc . ;
- b. Unit 4 Radioactivity- 1 1 ** <2.8x10 2 pCi' 28 High Gaseous #. cc (SPING).
or
<1.0x108 CPM -
k .(PRMS) 5 i5i 5 5
= . C 3#
5 C m F d "
. 4
.. . . . . ,. .. . . . . _ . . _ .., .- . , _ ._{ ___s_,
, ._ , , , . , . _ _ , , ,__ ., . , . , , , y ', , .,_
l TABLE 3.3-4 (Continued) l TABLE NOTATIONS i
- During CORE ALTERATIONS or movement of irradiated fuel within the ,
containment comply wit.) Specification 3/4.9.13. ! With irradiated fuel in the spent fuel pit . I
# Unit 4 Spent Fuel Pool Area is monitored by Piant Vent radioactivity instrumentation.
Note 1 Either the particulate or gaseous channel in the OPERABLE status 1 will satisfy this LCO. ! hote 2 Containment Gateous Monitor Setpoint = (3.2 x 109 CPM, ( F ) Actual Purge Flow j Where F = Design Purge Flow (35,000 CFM) ' Setpoint may vary according to current plant conditions provided I that the release rate does not exceed allowable limits provided in l Speci fication 3.11.2.1. ACTION STA1EMENTS l ACTION 26 - lii MODES 1 thru 4: With both the Particulate and Gaseous ! Radioactivity Monitoring Systems inoperable, operation may continue for up to 7 days provided:
- 1) A Containment sump level monitoring system is OPERABLE,.
L
- 2) Appropriate grab samples are obtained and analyzed at least once per 24 hours, l l
- 3) A Reactor Coolant System water inventory balance is .l performed at least once per 8 hours during steady state '
operation except when operating in shutdown cooling mode, and ,
- 4) Containment Purge, Exhaust and Instrument Air Bleed Valves are maintained closed. 1 1
Otherwise, be in at least HOT STANDBY within the next 6 hours i and in COLD SHUTDOWN within ee following 30 hours (ACTION 27 applies in MODES 5 and 6), l l l l I l l l
. TURKEY POINT - UNITS 3 & 4 3/4 3-37 AMENDMENT NOS.137AND 132 l
TABLE 3.3-4 (Continued) ACTION STATEMENTS (Continued) ACTION 27. ' In MODES 5 or 6 (except during CORE ALTERATION or movement of irradiated fuel within the containment): With the number of OPERABLE Channels less tnan the Minimum Channels OPiRABLE requirement perform the following: ;
- 1) Obtain and analyze appropriate grab samples at leart once per 24 hours, and i
- 2) Monitor containment atmosphere with area radiation monitors.
Otherwise, isolate all penetrations that provide direct access from the containment atmosphere to the outside atmosphere. During CORE ALTERATION or movement of irradiated fuel within the containment: With the number of OPERABLE Channels less than the Minimum Channels OPERABli requirements, comply with ACTION '
. statement requirements of Specification 3.9.9 and 3.9.13.
ACTION 28 - With the number of OPERABLE channels less than the Minimum , Channels OPERABLE requirement, immediately suspend operations ; in the Spent Fuel Pool area involving spent fuel manipulations. , l I i l 1 TURKEY POINT - UNITS 3 & 4 3/4 3-38 AMFNOMENT NOS.137AND 132
TABLE 4.3-3 ~
$E RADIATION' MONITORING INSTRUMENTATION'FOR PLANT Q -
OPERATIONS SURVEILLANCE REQUIRtritNTS 8 5 ANALOG
.- CHANNEL M00ES FOR WilCH c CHANNEL CHANNEL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECF CALIBRATION TEST IS REQUIRED m
w 1. Containment e a a. Containment Atmosphere S R M All Radiaoctivity--High
- b. RCS Leakage Detection
- 1) Particulate Radio- S R M 1, 2, 3, 4 activity M 2) Gaseous Radioactivity S R M 1,2,3,4
- 2. Spent Fuel Pool Areas
- a. Unit 3 Radioactivity--High Gaseous S R M *
> b. Unit 4 (Plant Vent)
M Radioactivity--High 5 Gaseous # E (SPING and PRMS) S R M
- 5 , .
2 TABLE NOTATIONS - 8 L
- With irradiated fuel in the fuel storage pool areas.
y # Unit 4 Spent Fuel. Pool Area is monitored by Plant Vent radioactivity instrumentation. 5 t; m 4 w sw, a 3 .a , " & -<w -- - - ,, -_ _
.i.w~, e, - - - , --r3 ._#__v _w_ -r____ u_.._._
a' _ - _ s,m_mu m_. ._ . _ .. _ _ . - . _ _.. _ _m._2._______ __m_. . . _ _ . -_ . ._ . . ,_ _ . , _ . _ _. _ . . ._m._
y >
, 1 i
I I L INSTRUMENTATION- I MOVABLE INCORE DETECTORS l RR LIMITING CONDITION FOR OPERATION I 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with.
- a. At least 16 detector thimbles when used for recalibration and check of the Excore Neutron Flux Detection System and monitoring the QUANDRANT POWER TILT RATIO *, and at least 38 detector thimbles I whenusedformonitoringFfg,F(Z)andFxy(Z).
9 l
~ b. A r.inimum of two detector thimbles per core quadrant, and
- c. !iufficient movable detectors, drive, and readout equipment to map these thimbles, j
APPLICABILITY: .When the Movable Incore Detection System is used for: ;
- a. Recalibration of the Excere Neutron Flux Detection System, or I
- b. Monitoring the QUADRANT POWER TILT RATIO *, or l c. Measurement of Ffg, q F (Z) and Fxy(Z).
ACTION: With the Movable Incore Detection System inoperable, do not use the system for l the above applicable monitoring or calibration functions. The provisions of
- j. Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS _ 4.3.3.2 The Hovable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours by normalizing each detector output when required for:
- a. Recalibration of the Excore Neutron Flux Detection System, or
- b. Monitoring the QUADRANT POWER TILT RATIO *, or -
- c. MeasurementofFfH,F(Z)andFxy(Z). ,
9 ,
- Exception to C.ie 16 detector thimble requirement of monitoring the QUADRANT l
POWER TILT RM IO is acceptable when performing Specification 4.2.4.2 usirg i two sets of tour symmetric thir.:bles, i l
, TURKEY POINT - UNITS 3 & 4 3/4 3-40 AMENDMENT NOS.137/.ND 132 L .
l
I INSTRUMENTATION , l ACCIDENT MONITORING INSTRUMENTATION : LIMITING CONDITION FOR OPERATION 3.3.3.3 The' accident monitoring instrumentation channels shown in Table 3.3-5 ! shall be OPERABLL' APPLICABILITY: As shown in Table 3.3-5. ! ACTION: i
- a. As shown in Table 3.3-5.
I
- b. The provisions of Specification 3.0.4 are not applicable to ACTIONS in Table 3.3-5 that reqtire a shutdown, t
i
. SURVEILLANCE REQUIREMENTS ,
4.3.3.3 Each accident monitoring instrumentation channel shall be demonstrated f OPERABLE by performance Of the CHANNEL CHECK and CHANNEL CALIBRATION at the frequencies shown in Table 4.3-4. ; e ln i i k l l l-l i TURKEY POINT - UNITS 3 & 4 3/4 3-41 AMENDMENT N05.137AND 132 1: , e l _ _ _- .. .
TABLE 3.3 ' d
!ll ACCIDENT MONITORING INSTRUMENTATION
,:, TOTAL MINIMUM APPLI-S NO. Of CHANNELS CABLE 5 INSTRUMENT CHANNELS OPERABLE M00ES ACTIONS h 1. Containment Pressure (Wide Range) 2 1 1,2,3 31, 32 i'i d 2. Containment Pressure (Narrow Range) 2 1 1,2,3 36
- p. 3. Reactor Coolant Outlet Temperature 2-2Detectok 1-2 Detectors 1, 2, 3 31, 32
. TH0T (Wide Range) per Channel per Channel
- 4. Reactor Coolant Inlet Temperature 2-2 Detectors 1-2 Detectors 1, 2, 3 31, 32 TCOLD (Wide Range) per Channel per Channel y 5. Reactor Coolant Pressure - Wide Range 2 1 1,2,3 31, 32 Y 6. Pressurizer Water Level 2 1 1, 2, 3 31, 32 0
- 7. Auxiliary Feedwater Flow Rate 2/ steam 1/stea; 1, 2, 3 31, 32 generator generato
- 8. Reactor Coolant System Subcooling Margin 2(2) 1(2) 1, 2, 3 31, 32 Monitor g 9. PORV Position Indicator (Primary Detector) 1/ valve 1/ valve 1,2,3 33
$ 10. PORV Block Valve Position Indicator 1/ valve 1/ valve 1, 2, 3 33 4
5 11. Safety Valve Position Indicator (Primary 1/ valve 1/ valve 1,2,3 32 i z Detector) 8
- 12. Containment Water Level (Narrow Range) 2 1 1,2,3 36 g 13. Containment Water Level (Wide Range) 2 1 1,2,3 31, 32 .
U f9 - e
'-_ _w _ _ _ +_ - , __w __-m -_ s_u.____._ __._...m..._mt- 2. _m_ - ._m _mm__...m.-___ _. . _ _ _ _ _ . _ . _ . _ _ , . _ . _ _
P-w TABLE 3.3-5 (Continued) ACCIDENT MONITORING INSTRUMENTATION , , m -TOTAL MINIMUM APPLI- ' 3 NO. OF Cl:ANNELS CABLE
'i INSTRUMENT CHANNELS OPERABLE MODES ACTIONS c- 14. In Core Thermocouples (Core Exit Thermo- 4/ core 2/ core 1, 2, 3 31, 32
( 15. couples) Containment High Range Area Radiation quadrant 2 1 1,2,3 34 [ o- 16. Reactor Vessel Level Monitoring 2(1) 1(1) 1, 2,.3 37, 38
- System
- 17. Neutron Flux, Backup NIS (Wide Range) 2 1 1,2,3 31, 32.
l: 18. Containment Hydrogen Monitors 2 1 1, 2 35 g 19. High Range-Noble Gas Effluent Monitors
- a. Plant Vent Exhaust 1 1 ALL 34 Y b. Unit 3-Spent Fuel Pit Exhaust 1 1 ALL 34 0 c. Cadenser Air Ejectors 1 1 1,2,3 34
- d. Main Steam Lines 1 1 1,2,3 34
- 20. RWST Water Level 2 1 1,2,3 31, 32
- 21. Steam Generator Water Level (Narrow Range) 2/sta. gen. 1/sta. gen. 1, 2, 3 31, 32 i
- 22. Containment Isolation Valve Position Indicatien* 1/ valve 1/ valve 1, 2, 3 39 E
y TABLE NOTATIONS x
, y 1. A channel is eight sensors in a probe. A channel is OPERABLE if a minimum of four sensors are OPERABLE.
s g 2. Inputs to this instrument are from instrument items 3, 4, 5 and 14 of this Table.
>
- Applicable for containment isolation valve position indication designated as pc accident monitoring instru-E mentation (containment. isolation valves which receive containment isolation Phase A, Phase B, or containment ventilation isolation signals).
~ u -s- g , ,- - .,=<___-_.-.-__a->u--. ._. _a_>._a. .A.,-.- - . . . - . _ ~ - - _ . . _ _ - -.
6 TABLE 3.3-5 (Continued) ACTION STATEMENTS ACTION 31 With the number of OPERABLE accident monitoring instrumentation channel (s) less than the Total Number of Channels either restore the inoperable channel (s) to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours. ACTION 32 With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE, either restore the inoperable channel (s) to OPERABLE status within 48 hours, or be in at least HOT STANDBY within the next 6 hours and in at least HOT SHUTDOWN within the following 6 hours. ACTION 33 Close the associated bicek valve and open its circuit breaker. ACTION 34 With the number'of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter (s), within 72 hours, and:
- 1) Either restore the inoperable channel (s) to OPERABLE status within 7 days of the event, or i
- 2) Prepare and submit a Special Report to the Commission L
pursuant to Specification 6.9.2 within 14 days.following the event outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status. ACTION 35 With one or both hydrogen monitor (s) inoperable, comply with Action Requirements of Specification 3.6.5. ACTION 36~ With the number of OPERABLE accident monitoring instrumentation , channels less than the Minimum Channel OPERABLE, either restore - the inoperable channel to OPERABLE status within 30 days, or be in at least HOT STANDBY within the next 6 hours and in at-least HOT SHUTDOWN within the following 6 hours. ACTION 37 With the number of OPERABLE channels one less than the Total Number of Channels, restore the system to OPERABLE status within 7 days. If repairs are not feasible without shutting down, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE scatus. - 4 TURKEY POINT - UNITS 3 & 4 3/4 3-44 AMENDMENT N05.137AND 132
4 L TABLE 3.3-5 (Continued) i ACTION STATEMENTS ) ACTION 38 With the number of OPERABLE channels less than the Minimum I ( ' Channels OPERABLE requirements, restore the inoperable channel (s) , ! to OPERABLE status within 48 hours. If repairs are not L feasible without shutting down
]
- 1. Initiate an alternate method of monitoring the reactor I vessel inventory;.and -l
-2. Prepare and submit a Special Report to the Commission l
! pursuant to Specification 6.9.2 within 30 days following : the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status; and
- 3. Restore at least one channel to OPERABLE status at the I next scheduled refueling. ,
4 Action 39 With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, verify position by an alternate . means (e.g. administrative controls, ERDADS, alternate position indication, or visual observation) within 2 hours, and restore the inoperable sbannel(s) within 72 hours or comply with-the , provisionsofSp..fication3.6.4foranInoperablecontainment isolation valve. i l i f TURKEY POINT - UNITS 3 & 4 3/4 3-45 AMENDMENT N05.137 AND1 32 L L , _
TABLE 4.3-4 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION
- 1. Containment Pressure (Wide Range) M R
- 2. Containment Pressure (Narrow Range) M R
- 3. Reactor Coolant Outlet Temperature - T M N H0T (Wide Range)
- 4. Reactor Coolant Inlet Tempercture - TCOLD M R (Wide Range)
- 5. ' Reactor Coolant Pressure - Wide Range M R
- 6. Pressurizer Water Level M R
- 7. . Auxiliary Feedwater Flow Rate M R
. 8. Reactor Coolant System Subcooling Margin Monitor M R l
- 9. PORV Position Indicator (Primary Detector) M R 10 PORV Block Valve Position Indicator M R
- 11. Safety Valve Position Indicator (Primary Detector) M R
- 12. Containment Water Level (Narrow Range) M R
- 13. Containment Water Level (Wide Range) M R
- 14. In Core Thermocouples (Core Exit Thermocouples) M R
- 15. Containment - High Range Area Radiation Monitor M R*
- 16. Reactor Vessel Level Monitoring System M R
- 17. . Neutron Flux, Backup NI3 (Wide Range) M R
- 18. Containment Hydrogen Monitor M R*
- 19. High Range - Noble Gas Effluent Monitors '
- a. Plant Vent Exhaust M R J
- b. Unit 3 - Spent Fuel Pit Exhaust M R
- c. CondenserAirEjectors M R d.. Main Steam Lines M R
- 20. RWST Water Level M R
- 21. Steam Generator Water Level (Narrow Range) M R
- 22. Containment Isolation Valve Position Indication M R
- Acceptable criteria for calibration are provided in Table II.F.1-3 of [
NUREG-0737. ' i TURKEY POINT -' UNITS 3 & 4 3/4 3-46 AMENDMENT N05.137AND 132
._. _ = _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . . , .
INSTRUMENTATION FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.4- As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-6 shall be OPERABLE.
, APPLICABILITY: Yhenever equipment protected by the fire detection instrument is required to be OPERABLE.
ACTION:
- a. With any, but not more than one-half the total in any fire zone, Function A fire detection instruments shown in Table 3.3-6 inoperable, restore the inoperable instrument (s) to OPERABLE status within 14 days or within the next I hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours (or monitor the containment air temperature at least once per haur at the locations listed in Specification 4.5.1.5).
- b. With more than one-half of the Function A fire detection instrument in any fire zone shown in Table 3.3-6 inoperable, or vIth any Function B fire detection instruments shown in Table 3.3-6 inoperable, or with any two or more adjacent fire detection instruments shown in Table 3.3-11 inoperable, within I hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours (or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5).
- c. With the fire watch patrol not established at the 18 foot level of the turbine area, restore the fire watch patrol within one hour, or prepare and submit a Special Report to the Commission.
- d. The provisions of Specifications 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.4.1 Each of the above required fire detection instruments which are accessible during plant operation shall be demonstrated OPERABLE at least once per 6 months by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Fire detectors which are not accessible during plant operation shall be demonstrated ' OPERABLE by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST during each COLD SHUTDOWN exceeding 24 hours unless performed in the previous 6 months. 4.3.3.4.2 The NFPA Standard 720 supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months.
. TURKEY POINT - UNITS 3 & 4 3/4 3-47 AMENDMENT N05.137AND 132
i, TABLE 3.3-6 , FIRE DETECTION INSTRUMENTS i FOR ESSENTIAL EQUIPMENT TOTAL NUMBER i INSTRUMENT LOCATION OF INSTRUMENTS . HEAT FLA C SMOKE FIRE ZONE AREA , (x/y)* Ci797 Ti797 4 - Aux. Bldg. Corrider E. 1)' (2/0) 5 - Chem. Drain / Laundry / Shower Tank Room (2/0) 9 - Laundry / Chemical Drain Tank Room (1/0) i 10 - Pipeway (11/0) 11 - Unit 3 RHR Heat Exchanger Room (5/0)*** 12 - RHR Pump 3A Room (2/0)*** 13 - RHR Pump 3B Room (2/0)*** i 14 - Unit 4 RHR Heat Exchanger Room (5/0)*** I 15 - RHR Pump 4A Room (2/0)*** 16 - RHR Pump 4B Room (2/0)*** t 19 - Unit 3 W Elect Pcoet Room (5/0)*** l
'20 - Unit 3 S Elect Penet Room (11/0) 21 - Instrument Shop (2/0) 22 - Radioactive Laboratory (2/0) I 26 - Unit 4 N Elect Penet Room (8/0) '
27 - Unit 4 W Elect Penet Room (6/0) 30 - Unit 4 Piping and Valve Room (4/0)*** 40 - Unit 3 Piping and Valve Room (4/0)*** 45 - Unit 4 Charging Pump Room (0/4) (3/0) 47 - Unit 4 Component Cooling Water Area (0/4) (5/2)*** 54 - Unit 3 Component Cooling Water Area (0/4) (4/2)*** 55 - Unit 3 Charging Pump Room (0/4) (3/0) 58 - Aux Bldg Corridor,_E1. 18' (18/0) 59'- Unit 4 Containment Electrical (10/0) Penet. Area *.* 60 - Unit 3 Containment Electrical (16/0) Penet. Area ** ! 61 - Reactor Control Rod Eqpmt Room - Unit 4 (4/0) 62 - Computer Room (11/0) 63 - Reactor Control Rod Eqpmt Room - Unit 3 (4/0) 67 - 4160V Switchgear 4B (10/0) - 68 - 4160V Switchgear 4A (6/0) 70 - 4160V.Switchgear 3B (10/0) 71 - 4160V Switchgear 3A (6/0) 72 - Emergency Diesel B (0/3 (1/0) (1/0) 73 - Emergency Diesel A (0/3 (1/0) (1/0) 74 - Emergency Day Tank Room B (1/1) 75 - Emergency Day Tank Room A (1/1) 76 - Unit 4 Turbine Lube Oil Reservoir (1/0) 79A- North-South Breezeway (0/6) (4/0) i 81 - Unit 4 Main Transformer (1/0) 82 - Unit 4 Aux Transformer Area (1/0) 84 - Unit 3 and 4 Aux Feedwater Pump Area (3/0) ;
-(DC Enclosure Bldg)
. TURKEY POINT - UNITS 3 & 4 3/4 3-48 AMENDMENT N05.137AND 132
TABLE 3.3-6 (Continued)
'[ FIRE DETECTION INSTRUMENTS FOR ESSENTIAL EQUIPMENT TOTAL NUMBER INSTRUMENT LOCATION OF INSTRUMENTS HEAT FLAME SMOKE FIRE ZONE ARE. (x/y)* Tx7y7 (x/y) 87 - Unit 3 Aux Transformer Area (1/0) 93 - 480V Load Center 4A and 4B (1/0) 94 - 480V Load Center 4C and 4D (2/0) 95 - 480V Load Center 3A and 3B (1/0) 96 - 480V Load Center 3C and 3D (2/0) 97 - Mechanical Equipment Room (1/0) 98 - Cable Spreading Room (16/15) 101 .RPI Inverter and MG Sets (1/0) 102- Battery Rack 4B (1/0) 103- Battery Rack 3A (1/0)
, 104- RPI Inverter and MG Sets (2/0) v 10f- i Control Room -
(1/0) (17/0) 103A- Train A Inverters (3/4) 1988- Train B Inverters (4/4)
- 109- Battery Rack 4A (1/0)
- 110- Battery Rack 3B (1/0)
, 113- Unit 4 Feedwater Platform (2/0)*** l 116- Unit 3 Feedwater Platform (2/0)*** 119- Unit 4 Intake Cooling Water Pump Area (4/0)*** 1 120- Unit 3 Intake Cooling Water Pump Area (4/0)*** i 132- Control Room Electrical Chase (1/0) N/A - 18' level of the Turbine Area (N/A)# (N/A)# (N/A)# l TABLE NOTAL l9NS l n (x/y): x is numoer of Function A (early warning fire dectection and notification only) instruments. l y is number of Function B (actuation of Fire Suppression Systems and , L early warning fire detection and notification) instruments. ~ .
- The fire detection instruments located within the containment are not required to be operable during the performance of Type A Containment '
Leakage Rate Test. nes
- Installed to meet the requirements of 10 CFR Part 50, Appendix R, Section III.G. ,
l
#: A fire watch patrol shall be established to inspect the 18 foot level I of the Turbine Area once each hour.
TURKEY POINT - UNITS 3 & 4 3/4 3-49 AMENDMENT NOS.137AND 132
o b 6 INSTRUMENTATION RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-7 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The Alarm / Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM). APPLICABILITY: At all times, except as indicated in Table 3.3-7. > ACTION:
- a. With a radioactive liquid effluent monitoring instrumentation channel Alarm / Trip Sotpoint less conservative than required by the above spe-cification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable, or change the setpoint so it is acceptably conservative,
- b. With less than the minimum number of radioactive liquid effluent :
l monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-7. Restore the inoperable instrumentation to OPERABLE-l status within 30 days and, if unsuccessful, explain in the next Semiannual Radioactive Effluent Release Report pursuant to Specifica-tion 6.9.1.4 why this inoperability was not corrected in a timely l- manner. l i c. The provisions of Specification 3.0.3 are not applicable. l SURVEILLANCE REQUIREMENTS l 4.3.3.5 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-5. l: l l TURKEY POINT - UNITS 3 & 4 3/4 3-50 AMENDMENT NOS.137A ND132
~
l 1 t TABLE 3.3-7
, RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE ACTION
~
- 1. Gross Radioactivity Monitors Providing Alarm and A*:tomatic Termination of Release
- a. Liquid Radwaste Effluent Line 1* 35 4
- b. Steam Generator Blowdown Effluent Line 1** 36
- 2. Flow Rate Measurement Devices ;
- a. Liquid Radwaste Effluent Line 1* 37
- b. Steam Generator Blowdown Effluent Line 1**/ Steam 37 generator i l
l I i l
. +
9
- Applicable during liquid effluent releases.
*' Applicable during blowdown operations.
TURKEY POINT - UNITS 3 & 4 3/4 3-51 AMENDMENT N05.137AND 132 3 e t
+ .v , . -- - - - -
r . TABLE 3.3-7 (Continued)- TABLE NOTATION ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided that prior to initiatin, a release:
- a. At least two independent samples are analyzed in accordance with Specification 4.11.1.1.1, and
- b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge line valving; Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, efflue'nt releases via
~
this pathway may continue provided grab samples are analyzed for gross (beta or gamma) radioactivity at a lower limit of detection of no more than 1 x 10 7 microcuries/ml or analyzed isotopically (Gamma) at a-lower limit of detection of at least S x 10 7 microcuries/ml:
- a. At least once per 12 hours when the specific activity of
, the secondary coolant is greater than 0.01 microcuries/ gram DOSE EQUIVALENT I-131, or-
- b. At least once per 24 hours when the specific activity of the secondary coolant is less than or equal to 0.01 micro-curies / gram DOSE EQUIVALENT I-131.
ACTION 37 - With the number of channels OPERABLE less than required by the - Minimum Channels OPERABLE requirement, efficent releases via this pathway may continue provided the flow rate is' estimated at least once per 4 hours during actual relear,es. Pump performance curves may be used to estimate flow. L i l 1
. TURKEY POINT.- UNITS 3 & 4 3/4 3-52 AMENDMENT NOS.137 AND 132 i
a TABLE 4.3-5 C DEI 0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS o ANALOG S CHANNEL 5 CHANNEL SOURCE CHANNEL OPERATIONAL e INSTRUMENT CHECK CHECK CALIBRATION TEST C -
% 1. Gross Radioactivity Monitors Providing m Alarm and Automatic Termination w of Release
[ a. Liquid Radwaste Effluents Line - D P R(2)* Q(1)
- b. Steam Generator Blowdown Effluent Line O M R(2) Q(1) -
- 2. Flow Rate Measurment Devices
{ a. Liquid Radwaste Effluent Line 0(3) N.A. R* Q
- b. Steam Generator Blowdown Effluent Lines D(3) N.A. R Q
- Channel calibration frequency shall be at least once per 18 months.
i TABLE NOTATIONS l (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and E control room alarm annunciation occurs if the instrument indicates measures levels above the Alarm / Trip E Setpoint. x E
* (2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the Nat 9nal Bureau of Standards (NBS) or using standards that have been obtained from suppliers that 5 participate in measurement assurance activities with NBS. These standards shall permit calibrating the P system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, U sources that 5cve been related to the initial calibration shall be used.
W 5 (3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK s shall be made at least unce per 24 hours on days on which continuous, periodic, or batch releases are M made. y . -, w e , g--.=~ , - _ _ _ _ _ - _ _ _ _ _ _ _ _ , _ _ _ _ _ _ _ _ _ _ _ _ _ , , _ _ _ , . _ , . , . . _ _ , , _ , _ _ , _ _ . . , _ , _
I INSTRUMENTATION l RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION ; l LIMITING CONDITION FOR OPERATION 3.3.3.6 The radioactive gaseous effluent monitoring instrumentation channels i shown in Table 3.3-8 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of Specification 3.11.2.1 and 3.11.2.5 are not exceeded. The Alarm / Trip Setpoints of these channels meeting Specification 3.11.2.1 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM. APPLICABILITY: As shown in Table 3.3-8 ACTION: , q
- a. With a radioactive gaseous effluent monitoring instru',sentation ,
l channel Alarm / Trip Setpoint less conservative than required by the ' above specification, immediately suspend the release of radioactive , gaseous effluents monitored by the affected channel, or declare the ; channel inoperable or change the setpoint so it is acceptably ; conservative.
- b. With less than the minimum number of radioactive gaseous effluent l monitoring instrumentation channels OPERABLE, take the ACTION shown ;
in Table 3.3-8. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful explain in the next l Semiannual Radioactive Effluent Release Report pursuant to Specifica- ; tion 6.9.1.4 why this inoperability was not corrected in a timely manner,
- c. The provisions of Specification 3.0.3 are not applicable.
t l SURVEILLANCE REQUIREMENTS 4.3.3.6 Each radioactive gaseous effluent monitoring instrumentation channel i shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the 1 frequencies shown in Table 4.3-6. l l l TURKEY POINT - UNITS 3 & 4 3/4 3-54 AMENDMENT N05.137AND 132
- ~ ~ -
.4g .a
~
. h TABLE 3.3-8 .,
C
$E RADI0 ACTIVE GASEOUS EFFLUENT M0WITORING INSTRUMENTATION R .
b ' MINIMUM CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION
',i c 1. GAS DECAY TANK SYSTEM 5
1 d a. Noble Gas Activity Monitor - l w Providing Alarm and Automatic 7 e. Temination of Release (Plant Vent Monitor) 1 45
- b. Effluent System Flow Rate Measuring Device 1
- 46
- 2. WASTE GAS DISPOSAL SYSTEM (Explosive Gas Monitoring System) i 5 a. Hydrogen and Oxygen Monitors 1 ** 49
- 3. Condenser Air Ejector Vent System
- a. Noble Gas Activity Monitor (SPING or PRMS) 1 # 47 l
g b. Iodine Sampler 1 N 48 m E 1
- c. Particulate Sampler 1 # 48 k d. Effluent System Flow Rate Measuring Device 1 M 46 fC e. Sampler Flow Rate Measuring Device 1 N 46 3;'
5 C fV 9 e p -- . ~ - , , , , - + 2.m,,- -,- - a,-m __ _ . __, , - - - . _ _ _ _ . -___________..___w_,__,,__n_. -
,,,__ , _ _ _ _ _ _ _ . . , , , _ , _ , , _ _ _ , , ,.a....,_... , _,a ,
TABLE 3.3-8 (Continued)- N RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUNENTATION 7 . _ r S
~'
MINIMUM CHANNELS t
'i INSTRUMENT OPERA 8LE APPLICABILITY ACTION l
c- 4. Plant Vent System (Include Unit ?'s 5 Spent Fuel Pool) m w a. Noble Gas Activity Monitor (SPING 1
- 47
- e. or PRMS) s
- b. Iodine Sampler 1
- 48 t
- c. Particulate Sampler 1
- 48 R*
- d. Effluent System Flow Rate Measuring 1
- 46 Device T
E e. Sampler Flow Rate Measuring Device 1
- 46
- 5. Unit 3 Spent Fuel Pit Building Vent
- a. Noble Gas Activity Monitor 1
- 47
- b. Iodine Sampler 1
- 48 2
i5i c. Particulate Sampler 1
- 48 z
k*
~
'd. Sampler Flow Rate Measuring Device 1
- 46
- 5 m
o E m . O 6 m
'*= -V v' _ W __4- _ "' "e "D _w 7L._ W*^'er_t. __ _T-' - e-L. _ * - - _ . _ _ _ -_.v_" m. _____d_ _ _.oe-ma . m__--.t. _ _. _e__..-u u- a,a,-_.a___,__.-__m._._.-__.2_m_ .___.. m__-_
r s TABLE 3.3-8 (Continued) 1 TABLE NOTATION ;
- At all times. ,
** During GAS DECAY TANK SYSTEM operation. l
# Applies during MODE 1, 2, 3 and 4. j
- N Applies' during MODE 1, 2, 3 and 4 when primary to secondary leakage is I detected as indicated by condenser air ejector noble gas activity monitor.. -l ACTION 45 - With the number of channels OPERABLE less than required by the ;
Minimum Channels OPERABLE requirement, the contents of the i tank (s) may be released to the environment provided that prior to initiating the release: ]
- a. At least two independent samples of the tank's contents -
are analyzed, and
- b. At least two technically qualified members of the facility ;
staff independently verify the relenc rate c31culations and discharge valve lineup l Otherwise, suspend release of radioactive effluents "ia this j pathway. ; ACTION 46 - With the number of channels OPERABLE less than requires by the . Minimum Channels OPERABLE requirement, effluent releases via this pathway may continua provided the flow rate is estimated at least once per 4 hours. ACTION 47 - With the number of channels OPERABLE less than required by the l Minimum Channels OPERABLE requirement, effluent releases via
- this pathway may continue provided grab samples are taken at least once per 12 hours and these samples are analyzed for radioactivity within 24 hours.
ACTI0n - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as , required in Table 4.11-2 and analyzed at least weekly. ACTION 49 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, operation of the GAS DECAY TANK SYSTEM may continue provided that grab samples are collected and analyzed for hydrogen and oxygen concentration at ' least a) once per 8 hours during degassing operations, and b) once per day during other operations. - 132 TURKEY POINT - UNITS 3 & 4 3/4 3-57 AMENDMENT NOS.137AND
p . .
~ o. _ ,
=' .
[~~
-TABLE'4.3-6
$E RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Q
S ANALOG 5 CHANNEL M00ES FOR WHICH
^
. CHANNEL SOURCE- CHANNEL OPERATIONAL SURVEILLANCE IS c- INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED 5
l- d 1. GAS DECAY TANK SYSTEM
- a. Noble Gas Activity Monitor - ,
a Providing Alarm and Automatic Termination of Release (Plant Vent Monitor) P P R(3) Q(1)
- b. Effluent System Flow Rate y Measuring Device P M.A. R N.A.
- Y 2. GAS DECAY TANK SYSTEM (Explosive
$ Gas Monitoring System)
- a. Hydrogen and Oxygen Monitors D N.A. Q(4,5) 'M **
- 3. Condenser Air Ejector Vent System
- a. Noble Gas Activity Monitor (SPING or PRMS) D. M R(3) Q(2) #
5 b. Iodine Sampler W N.A. N.A. N.A. M M 5 c. Particulate Sampler W 'N.A. N.A. N.A. N z - 8 d. Effluent System Flow Rate
~_ Measuring Device D N.A. R N.A. N O '
5 i::
~
s
, . - + ,
y TABLE 4.3-6 (Continued)
. e- -
RADI0 ACTIVE GASEOUS EFFLUENT NONITORING INSTRUMENTATION SURVEILLANCE REQUIREENTS o 8 ANALOG
*i CHANNEL MODES FOR WHICH e
CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE 15 g INSTRUMENT CHECK CI'M CALIBRATION TEST REQUIRED d 3. Condenser Air Ejector Vent System w (Continued) e. a e. Samp16 Flow Rate Measuring Devite D N.'A. R N.A. N-
- 4. Plant Vent System (Include Unit 4's Spent Fuel Pool)
- a. Noble Gas Activity Monitor j
{ w (SPING or PRMS) .D M (3,6) Q(2) l b. Iodine Sampler W N.A. N.A. N.A. *
- c. Particulate Sampler W N.A. N.A. N.A.
- l' l d. Effluent System Flow Rate Measuring Device
- L D N.A. (6) N.A.
E m e. Sampler Flow Rate Measuring 5 Device D N.A. (6) N.A.
- i3i
'i l 5 r
U 32 5 l C
~
l
. = . - _ . - ._ _ _ _ - -
. _ _ _ _ , _ .. . - - _.__.m._________m. _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ . . _ . . _ . . _ . _ _ _ _ . _ . _ _ _ _ _
a n.1 TABLE 4.3-6 (Continued) R RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS L Q l v l S ANALOG l 5 CHANNEL MODES FOR WHICH
. CHANNEL SOURCE CHANNEL' OPERATIONAL SURVEILLANCE IS-c- INSTRUMENT CHECK CHECK CALIt1 RATION TEST REQUIRED l 5 d 5. Unit 3 Spent Fuel Pit Building Vent l- w l e. a. Noble Gas Activity Monitor D M R(3) Q(2) s
- b. Iodine Sampler W N.A. N.A. N.A. *
- c. Particulate Sampler W N.A. N.A. N.A.
- t' d. Sampler Flow Rate Measuring Device D N.A. R N.A.
- w h
TABLE NOTATION At all times. During GAS DECAY TANK SYSTEN operation.
# Applies'during MODE 1, 2, 3 and 4.
## Applies during MODE 1, 2, 3 and 4 when primary to secondary leakage is detected as indicated by
> condenser air ejector-noble gas activity monitor. .
4 4 5 (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and 5 control room alarm annunciation occurs if the instrument indica,tes measured levels above the Alare/ Trip 5 Setpoint.
~
(2) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that if the instrument indicates measured levels above the Alarm Setpoint, alarm annunciation occurs in the control room (for PRMS only) and in ti the computer room (for SPING only). N o (3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified C by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that
~
participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over it; intended range of energy and measurement range. .For subsecuent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
-s,, _g-.--_ - - . - -_i-- _- . . - _ _ _ - - _ _ ~ ---
l i 1 TABLE 4.3-6 (Continued) L TABLE NOTATIONS (Continued) (4) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal.
- a. - One volume percent hydrogen, balance nitrogen, and
- b. Four volume percent hydrogen, balance nitrogen.
(5) The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:
- a. One volume percent oxygen, balance nitrogen, and
- b. Four volume percent oxygen, balance nitrogen.
(6) CHANNEL CALIBRATION frequency shall be at least once per 18 months. a J i TURKEY POINT - UNITS 3 & 4 3/4 3-61 AMENDMENT N05.137AND 132 l- s
I f.
'3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION d
STARTUP'AND POWER OPERATION LIMITING' CONDITION FOR OPERATION
., t (l
.j{ 3.4.1.1 All reactor coolant loops shall be in operation.
APPLICABILITY: MODES 1 and 2.
~
ACTION: With less than the above required reactor coolant loops in operation, be in at least HOT STANDBY within 6 hours. i SUREVEILLANCE REO,UIREMENTS o l 4.4.1.1 The above required reactor coolant loops shall be verified in opera-tion and circulating reactor coolant at least once per 12 hours. f
.I TURKEY POINT - UNITS 3 & 4 3/4 4-1 AMENDMENT NOS.137 AND 132 G .
e
REACTOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 All of the reactor coolant loops listed below shall be OPERABLE with all reactor coolant loops in operation when the Reactor Trip breakers are closed and two reactor coolant loops listed below shall be OPERABLE with at ! least one reactor coolant loop in operation when the Reactor Trip breakers are open: *
- a. Reactor Coolant Loop A and its associated steam generator and reactor l coolant pump, j
- b. Reactor Coolant loop B and its associated steam generator and reactor coolant pump, and ]
- c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump.
]
APPLICABILITY: MODE 3 ACTION: a.' With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours or be i in HOT SHUTDOWN within the next 12 hours. I
- b. With less than three reactor coolant loop in operation and the Reactor Trip breakers in the closed position, within I hour open the Reactor Trip breakers. ,
- c. With no reactor coolant loop in operation suspend all operations involvingareductioninbore,concentratIonoftheReactorCoolant i l System and immediately initiate corrective action to return the required reactor coolant loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reacar coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability. 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verify-ing_ secondary side water level to be greater than or equal to 10% at h st once l per 12 hours. 4.4.1.2.3 The required reactor coolant loops shall be verified in operation
- and circulating reactor coolant at least once per 12 hours.
*All reactor coolant pumps may be deenergized for up to I hour provided:
(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation temperature. TURKEY POINT - UNITS 3 & 4 3/4 4-2 AMENDMENT NOS.137AND 132 1
I a i REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION _ 3.4.1.3 At least two of the loops listed below shall be OPERABLE and at least one of.these loops shall be in operation:*
- a. Reactor Coolant Loop A and its associated stear, generator and reactor coolant pump,**
- b. Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,**
- c. Reactor Coolant Loop C and its associated steam generator and reactor coolant pump,**
- d. RHR Loop A, and
- e. RHR Loop B.
APPLICABILITY: MODE 4. ! ACTION:
- a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is an RHR loop, be in COLD SHUTDOWN within 24 hours.
- b. With no loop in operation, suspend all operations. involving a reduc-tion in boron concentration of the Reactor Coolant System and -1 immediately initiate corrective action to return the required loop to operation.
I
*All reactor coolant pumps and RHR pumps may be d a nergized for up to 1 hour ,
provided: (1) no operations are permitted that would cause dilution of the . Reactor Coolant System boron concentration, and (2) core outlet temperature 1 is maintained at least 10 F below saturation temperature.
**A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 275 F unless the secondary water temperature of each steam generator is less than 50 F above ;
each of the Reactor Coolant System cold leg temperatures. l TURKEY POINT - UNITS 3 & 4 3/4 4-3 AMENDMENT NOS.137AND 132
p -- 1
' REACTOR COOLANT' SYSTEM SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required reaciar coolant punip(s), if not in operation, shall be i determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 10% at least once per 12 hours. 4.4.1.3.3 At least one reactor coolant or RHR loop shall be verified in operation and circulating nactor coolant at least once per 12 hours. l l i 1 l j i
.. j e
-l l
I
)
TURKEY POINT - UNITS 3 & 4 3/4 4-4 AMENDMENT NOS.137AND 132 l 1
- k
\
REACTOR-COOLANT SYSTEM
'* COLO SHUTDOWN
- LOOPS-FILLED LIMITING CONDITION FOR OPERATION
~ 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and u- Lin operation *, and either: .
- a. One additional RHR loop shall be'0PERABLE**, or
- b. The secondary side water level of at _least two steam generators shall be greater than 10%.
APPLICABILITY: MODE 5 with reactor coolant loops filled ***.
> ACTI0N:
- a. With one-of.the 'RHR loops inoperable or with les:, than the required steam generator water level, immediately initi.*.te corrective action to return the inoperable RHR loop to OPERABLE status or restore the required steam generator water level as soon as possible.
- b. With no RHR loop in operation, suspend all opecations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.'4.1.1 The secondary side water level of at least two steam ger.erators when required shall be= determined to be within. limits at-least once per 12 hours.
. 4.4.1.4.1.2 At least one RHR loop shall be determined ta be in operation and circulating reactor coolant at least once par 12 hours.
*The RHR pump may oe deenergized for up to 1 hour provided: (1) no opera-tions are permitted that would cause dilution of the P.aactor Coolant System
, boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
**0ne KHR loop may be inoperable for up to 2 hours for surveillance testing provided the other "HR loop is OPERABLE.
***A reactor coolant pump shall not be started with one or more of the Reactor Coclant System ccid leg temperatures less than or equal to 275 F unless the secondary water temperature of each steam generator is less than 50 F above I each of the Reactor Coolant System cold leg temperatures. J I
TURKEY POINT - UNITS 3 & 4 3/4 4-5 AMENDMENT NOS.137AND 132
0 REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS NOT FILLE 0 _ LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two residual heat removal (RHR) loops shall be OPERABLE
- and at least one RHR loop.shall be in operation.**
APPLICABILITY: : MODE 5 with reactor coolant loops not filled. ACTION:
- a. With less than the above required RHR loops OPERABLE, immediately initiate corrective action to return the required RHR loops to OPERABLE status as soon as possible.
- b. With no' RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reacte Coolant System and immediately initiate corrective action to n ta:n the required RHR loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.4.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at .least once per 12 hours.
*0ne RHR loop may be inoperable for up to 2 hours for surveillance testing provided the other RHR loop is OPFRAEi.E.
~
**The RHR pump may be deanergized for up to 1 hour provided: (1) no opera-tions are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
TURKEY POINT - UNITS 3 & 4 3/4 4-6 AMENDMENT N05.137AND 132
~
Nw m
,, lha? m -
y l lg . > REACTOR COOLANT SYSTEM-- 3/4.4.2 SAFETY VALVES
, 9J4 SHUTDOWN.
LIMITING CONDITION FOR OPERATION Al 3 4.2.1 '. minimum of one pressurizer Code safety valve shall be OPERABLE
- with a lift setting of 2485 psig i 1%.**
APPLICABILITY: -MODES 4 and 5. e. av ACTION:
, n;s t With no_ pressurizer Code' safety valve OPERABLE, immediately suspend all-opera-
.tions involving positive reactivity changes and place an OPERABLE RHR' loop 46 , j -into operation in the shutdown cooling mode.
N .
, n.
p i b
' SURVEILLANCE REQUIREMENTS ,
n; Ja g 4.4.2.1 -No additional requirements other than those required by
, Specification-4.0.5.
e
/ .
4
'*While in MODE 5, an equivalent size vent pathway may be used provided that the. vent pathway is not isolated or sealed.
**The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
ii TURKEY POINT - UNITS 3 & 4 3/4 4-7 AMENDMENT NOS.137AND 132 ( 6 i . . . . . ,
p _ ,
-T
![ REACTOR COOLANT SYSTEM y w
- 0PERATING , a t i LIMITING CONDITION FOR OPERATION
'3.4.2.2 All' pressurizer Code safety valves shall be OPERABLE with a lift' setting of 2485 psig i 1%.*
APPLICABILITY: MODES 1 2 and 3. ACTION:-
- With _one pressurizer Code safety valve inoperable, either restore the inoper -
able valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY'
-within 6 hours and-in at least HOT SHUTDOWN within.the following 6 hours.
. SURVEILLANCE REQUIREMENTS ,
4.4.2.2 No additional requirements other than those required by Specification 4.0.5.
\
l E 1 i-i.
*The li f t' setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
r
.k a TURKEY POINT - UNITS 3 & 4 3/4 4-8 AMENDMENT N05.137AND 132
*+'s t - *
'l REACTOR COOLANT SYSTEM 3/4.4.3- PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3 The pressurizer shall'be OPERABLF. with a water volume of less than or -
equal to 92% of indicated level,.and at.least two groups of pressurizer heaters each having a capacity of at least 125 kW and capable of being , supplied by emergency. power. APPLICABILITY: MODES 1, 2, and 3. ACTION: a.- With only oaa qroup of pressurizer heaters OPERABLE, restore at least two grour - '7RABLE status witnin 72 hours or be in at.least HOT STANDBY next 6 hours and in HOT SHUTDOWN within the follow-ing 6 he
- b. With the pressurizer otherwise incpe:'ble, be in at least HOT STANDBY with the Reactor Trip System breako . pen within 6 hours and in HOT SHUTDOWN within the following 6hourt SURVEILLANCE REQUIREMENTS, 4.4.3.1 TheJpressurizer water volume shall oe determined to be within its limit at least^once per 12 hours.
4.4.3.2' The capacity of each of the above required groups of pressurizer heaters shall be verified by energizing the heaters and mea'suring circuit current at least once per 92 days. 4.4.3.3'. The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at-least once per 18 months by demonstrating the capabil-ity to power the heaters from the emergency power. TURKEY POINT - UNITS 3 & 4 3/4 4-0 AMENDMENT NOS.137 AND 132 t
n.
'h:
- REACTOR COOLANT SYSTEM 3/4.4.4 - RELIEF VALVES LIMITING CONDITION FOR OPERATION _
3.4.4 Each power-operated relief valve (PORV) block valve shall be OPERABLE.
' APPLICABILITY: _ MODES 1, 2 and 3.
ACTION:
- a. With one or more block valve (s) inoperab'le, within 1 hour either
_ restore the block valve (s) to OPERABLE status, or close the block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours and HOT. SHUTDOWN within the following 6 hours, b. The provisions of Specification 3.0.4 are not applicable, l l SURVEI'LLANCE REQUIREMENTS l: 1 4.4.4 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel , unless.the block valve is closed with power removed in order to meet the requirements of Specification 3.4.4 or is closed to provide an' isolation function. 1 J t-1 1. L TURKEY POINT - UNITS 3 & 4 3/4 4-10 AMENDMENT N05.137AND 132 l F
5 y ). i( [ j y : REACTOR COOLANT' SYSTEM N - 3/4.4.5' STEAM GENERATORS -f 3
. LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3.and 4.
- ACTION:
With one or more steam generators inoperable, rest' ore the inoperable genera- '
. tor (s) to OPERABLE status prior to increasing T,yg above 200 F.
e p , SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of
. Specification 4.0.5.
4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator U shall.be determined OPERABLE during shutdown by selecting and inspecting at-least the minimum number of steam generators specified in Table 4.4-1, 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the E corresponding action required shall be as specified in Table 4.4-2. The l- inservice inspection of steam generator tubes'shall be performed at the fre-L quencies. specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. Th'e - tubes se.lected for.each-inservice inspection shall include at least 3% of the total-number of tubes in all steam generators; the tubes selected for these I inspections shall be selected on a random basis except: r
- a. Where experience in similar plants with similar water chemistry l indicates critical areas to be inspected, then at least 50% of the i tubes inspected shall be from these critical areas;
- b. The first sample of tubes selected for each inservice inspection .
(subsequent to the preservice inspection) of each steam generator shall include: l i. , i TURKEY POINT - UNITS 3 & 4 3/4 4-11 AMENDMENT N05.137AND 132
4 L REACTOR ~ COOLANT SYSTEM
'y, ,
- STEAM GENERATORS-SURVEILLANCE REQUIREMENTS (Continued)
. 1) Allnonpluggedtubes/thatpreviouslyhaddetectablewallpene-y trations (greater than 20%),
- 2) Tubes in those areas where experience has indicated potential problems, and
- 3) A tube inspection.(pursuant to Specification 4.4.5.4a.8) shall be performed on each selected tube. If any selected tube does not permit the' passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected and subjected to a tube inspection.
- c. .The tubes selected as the second and third samples in the inservice inspection may be less than a full tube inspection by concentrating (selecting at least 50% of the tubes to be inspected) the inspection on those areas of the tube sheet array and on those poM lons of the tubes where tubes with imperfections were previously ft ad.
The' results of ea-h sample inspection shall be classified into one of the - following three ca'.egories: Category ' Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes;and none of the inspected tubes' are defective. C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes. C-3 More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective. Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations. TURKEY POINT - UNITS 3 & 4 3/4 4-12 AMENDMENT NOS.137AND 132
d .t-s i 8
, REACTOR'C00LANT SYSTEM:
e STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued)
'/
4.4,5.3 Inspaction Frequencies - The above required inservice inspections of
, steam generator tubes shall be performed at the following frequencies:
- a. The first' inservice inspection shall be performed after 6 Effective
' Full Power. Months but within 24 calendar months following replace -
; ment of: steam generators. Subsequent inservice inspections shall be-performed at intervals of not ~less than 12 nor more than 24 calendar months'after the presious inspection. If two consecutive inspections following. service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradatjon has not continued and no additional degradation has occurred, the inspeccion interval may be extended to a maximum of -
once per 40 months.-
- b. If the results of the inservice inspection of a steam generator con-ducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least '
once per 20 months. The increase in inspection frequency shall apply-until- the subsequent inspections satisfy the criteria of Specifica-tion 4.4.5.3a; the interval may then be extended to a maximum of once per 40 months; and Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:
- 1) Primary-to-secondary tubes leak (not including leaks originating from tube-to-tube sheet welds)-in excess of the limits of Specification 3.4.6.2, or
- 2) A seismic occurrence greater than the Operating Basis Earthquake,,
or-
- 3) A loss-of-coolant accident resulting in rapid depressurization of the primary system, or
- 4) A main steam line or feedwater line break resulting in rapid depressurization of the a'fected steam generator. .
l TURKEY POINT - UNITS 3 & 4 3/4 4-13 AMENDMENT NOS.137AND 132
t REACTOR COOLANT--S STEM
-STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.4 Acceptance criteria
- a. As used in this specification:
- 1) Imperfection means an exception to the dimensions, finish-or contour of a tube from that re. quired by fabrication drawings or specifications. Eddy-current testing indications below'20% of ~
the nominal tube wall thickne.ss, if detectable, may be con-sidered as imperfections;
- 2)
Dearadation means a servico-induced cracking,
wastage, wear or general corrosion occurring on either inside or outside of a tube;
- 3) Degraded Tube means a tube containing imperfections greater than or equal to 20% of the. nominal wall thickness caused by degradation;
- 4) % Dearadation means'the percentage of the tube wall thickness affected .or r emoved by degradation;.
- 5) Defect means an imperfection of such severity that it exceeds the plugging limit. A tube-containing a' defect is defective.
,6) Pluggina Limit means the imperfection depth at or beyond which the tube shall be removed from service because it may become enserviceable prior to the next inspection and is equal to 40%
of the nominal tube wall thickness;
- 7) Unserviceable describes the condition of a tube.if it leaks or contains a defect large enough to affect.its structural integ-rity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line or feedwater line break as specified in 4.4.5.3c., above;
- 8) Tube Inspection means an irispection of the steam generator tube from the point of entry (hot leg side) completely around the U-t end to the top support of the cold leg, or from the point of entry (cold leg side) completely around the U-bend and to the bottom of the hot leg; and TURKEY POINT - UNITS 3 & 4 3/4 4-14 AMENDMENT NOS.137AND1132
E' k 3
'\
REACTOR COOLANT' SYSTEM
. STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) 6
- 9) Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing.
- b. The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging
- limit and all tubes containing through-wall cracks) required by Table 4.4-2. -
4.4.5.5 Reports
- a. .Within 15 days following the completion of each inservige inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2;
.b. The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a Special Report pursuant to-Specification 6.3.2 within 12 months following-the completion of the inspection. This Special Report shall include:
1). Number and extent of tubes inspected,
.2) Location and percent of wall-thickness penetration for each indication of an imperfection, and
- 3) Identification of tubes plugged,
- c. Results of steam generator tube. inspections which fall into Category-
-C-3 shall be reported to the Commission pursuant to 10'CFR Part 50.72 and prior to resumption of plant operation. -This report shall provide a description of investigations conducted to determine'cause of the tube degradation and corrective measures taken to prevent recurrence.
l
' TURKEY POINT - UNITS 3 & 4 3/4 4-15 AMENDMENT NOS.137AND 132
TABLE ~4.4-1 . MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION Preservice Inspection No Yes No. of Steam Generators per Unit Three Three-First Inservice Inspection All Two ; Second & Subsequent Inservice Inspectionc- Onel One?
- Table Notation
- 1. The inservice inspection may be. limited to one steam generator on a rotat-ing schedule encompassing M of the tubes if the results of the first or previous inspections indicate that all steam generators are performing in '
a like manner. -Note that under some circumstances, the operating condi-tions in one or more steam generators may be found to be more severe than those in other steam generators. Under such circumstances the sample sequence shall be modified to inspect the most severe conditions.
- 2. :The other steam generator not inspected during the first inservice inspec-tion shall be inspected. The third and subsequent inspections should follow the instruction described in 1 above, d
a 4 e f TURKEY POINT - UNITS 3 & 4 3/4 4-16 AMENDMENT N05.137AND 132
TABLE 4.4-2 ~ STEAM GENERATOR TUBE-INSPECTION hk SAMPLE. -1st. SAMPLE INSPECTION 2nd SAMPLE INSPECTION 3rd SAMPLE INSPECTION El SIZE i ,3 Result Action Required Result Action Required Result Action Required 3 A minimum C-1 None N/A N/A' N/A N/A ! c: of S Tubes 25 per S.G. C-2 Plug defective tubes C-1 None N/A N/A-51 and inspect addi-co tional125 tubes in C-2 Plug defective tubes C-1 .None
- e. this S.G. inspect additional 45 4, tubes in this S.G. C-2 Plug' defective
-tubes-C-3 Perform action for-C-3 result ti :of first sample z, -
? C-3 Perform action for N/A N/A t!. C-3 result of first sample C-3 Inspect all tubes All other None N/A N/A in this S.G. plug S.G.s are defective tubes and C-1 inspect 25 tubes in 3, each other S.G. Some S.G.s Perform action for N/A N/A Ri C-2 but no C-2 result of second is Notification to NRC additional sample
#i pursuant to Sec- S.G.s are 25 tion 4.4.5.Sc. C-3 h! Additional Insrect all tubes'in N/A N/
5.G. is _m each S.G. and plug E3 C-3 defective tubes. 3: Notification to NRC t' pursuant to Section t; 4.4.5.5C.
~
S=f%Wherenisthenumberofsteamgeneratorsinspectedduringaninspection. m _. ___ _ _" _ _ _ _ _ _ _ _ _ _ _ _ .- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ' _ _ _ . _ _ _ _ _ _ . .
- _ - ~. . .mwn-
i REACTOR-COOLANT SYSTEM
};'4. 4 ~. 6 REACTOR COOLANT SYSTEM LEAKAGE 1
{AKAGEDETECTIONSYSTEMS LIMITING CC':DITION FOR OPERATION 3.4.6.1 1The following Reactor Coolant System Leakage Detection Systems shall ) be OPERABLE: 'i
- a. The Containment Atmosph'ere Gaseous or Particulate Radicactivity i Monitoring System, and , .. j
- b. A Containment Sump Level Monitoring System. j APPLICABILITY: MODES 1, 2, 3 and 4. l ACTION:-
- a. W'ith botn the Particulate and Gaseous' Radioactivity Monitoring Systems inoperable, operation may continue for up to 7 days provided:
-1) A Containment Sump Level Monitoring System is OPERABLE; 1
- 2) . Appropriate grab samples are obtained and analyzed at least once per 24 hours; 1
L 3) A Reactor Coolant System water inventory balance is performed l at least'once per 8 hours during steady state operation except ; when operating in shutdown cooling mode; and .J
- 4) Containment Purge, Exhaust and Instrument Air Bleed valves-are J maintained closed. .l
.0therwise, be in at'least HOT STANDBY within the next 6 hours and in i COLD SHUTDOWN within the following 30 hours. l l
b .- With no Containment Sump Level Monitor'ing System operable, restore at least one Containment Sump Level Monitoring System to OPERABLE , i status within 7 days, or be in at least HOT STANDBY within 6 hours ! and in COLD SHUTOOWN within the following 30 hours. .l l SURVEILLANCE REQUIREMENTS i 9 l L 4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by: l
- a. Containment Atm3 sphere Gaseous and Particulate Monitoring Syste'-
performance of CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG ',HANNEL OPERATIONAL TEST at the frequencies specified in Table 4.3 .5, and i L b. Containment Sump Level Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months. h TURKEY POINT - UNITS 3 & 4 3/4 4-18 AMENDMENT NOS.137AND 132
REACTOR C00 LAND SYSTEM-OPERATIONALLEAKiGlE - 4 LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant. System leakage shall be limited to:
- a. No PRESSURE BOCNDARY LEAKAGE,
- b. 1 GPM UNIDENTIFIED Lt W F ,
- c. 1 GPM total primary-to-secondary leakage through all steam generators '
and 500 gallons per day through any one steam generator,
- d. 10 GPM IDENTIFIED LEAKAGE from.the Reactor Coolant System, and
- e. leakage as specified in Table 3.4-1 up to-a maximum of 5 GPM at a Reactor Coolant-System pressure of 2235 1 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.*
APPLICABILITY: MODES _1, 2, 3 and 4 ACTION:
- a. With any PRESSURE B0UNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours,
- b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, and leakage from Reactor Coolant System Pressure: Isolation Valves, reduce the leakage rate to within limits within 4 hours or be in at least HOT STANDBY
- within the next 6 hours and in C0l0 SHUTDOWN within the following 30 hours.
- c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than allowed by 3.4.6.2.e'above operation may continue provided:
- 1. W'. thin 4 hours verify that at least two valves in each high pressure line having a non-functional valve are in,'and remain in that mode corresponding to the isolated condition, i.e.,
manual valves shall be locked in the closed position; motor operated valves shall be placed in the closed position and power supplies deenergized. Follow applicable ACTION statement for the a.ffected system, and
- Test pressures less than 2235 psig are allowed. Minimum differential test pressure shall not be less than 150 psid. Observed leakage shall be adjusted for the actual test pressure up to 2235 psig assuming the leakage to be directly proportional to pressure differential to the one-half power.
TURKEY P0 INT - UNITS 3 & 4 3/4 4-19 AMENDMENT NOS.137AND 132
.V
4 l ),* s REACTOR-COOLANT. SYSTEM
-OPERATIONAL LEAKAGE !
LIMITING CONDITION FOR OPERATION (Continued) i E I
- 2. The leakage
- from the remaining isolation valves in each high l '. . pressure line having a valve not meeting the criteria of L Table 3.4-1, as listed in Table 3.4-1, shall be determined and j' recorded daily. The positions of the other valves located in the high pressure line having the leaking valve shall be , !
recorded daily unless they are manual valve inside L containment. ,U Otherwise be in at least HOT STANDBY within the next 6 hours and in L COLD SHUTDOWN within the following 30 hours,
- d. With any Reactor Coolant System Pressure Isolation Valve leakage l ' greater than 5 gpm, reduce leakage to below 5 gpm within I hour, or 4 be in at least HOT STANDBY within the next 6 hours and in COLD SHUT-l- DOWN within the following 30 hours. j 1
l SURVEILLANCE REQUIREMENTS l l 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within. l l; each of the'above limits by:
- a. Monitoring the containment atmosphere gaseous or particulate radio- '
i activity monitor at least once per 12 hours. < E
- b. Monitoring the' containment sump level at least once per 12 hours.
' c; Performance' of a Reactor Coo. ant System wate'r inventory balance l within 12 hours after achieving steady-state operation ** and at least once per 24 hours thereafter. during steady-state operation, except-that not more than 48 hours shall elapse between any two successive inventory balancec; and
- d. Monitoring the Reactor Head Flange Leakoff System at least once per-I' 24 hours.
4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve-specified in l Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage
- to be within l its limit:
- a. At least once per 18 months.
- b. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN.
for 72 hours or more and if leakage testing has not been performed in the previous 9 months, and *
- c. -Prior to returning the valve to service following maintenance, repair or replacement work on the valve. -
"To satisfy ALARA requirements, leakage may be measured indirectly (as from ,
! the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
**RCS average coolant temperature being changed by less than 5*F/ hour.
TURKEY POINT - UNITS 3 & 4 3/4 4-20 AMENDMENT NOS 137AND 132
-REACTOR COOLANT-SYSTEM-OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION (Continued)
'd. Following valve actuation due to automatic or manual action or flow
-through-the valve:
- 1. Within 24 hours by, verifying valve closure, and
- 2. Prior to entering Mode 2 by verifying leakage rate.
The provisions of Specification 4.0.4 are not applicable for entry into. MODE 3 or 4.- 1 1 . t
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TURKEY POINT - UNITS 3 & 4 3/4 4-21 AMENDMENT NOS.137AND 132
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t ' c:; 1 ' TABLE'3.4-1
- REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES t VALVE NUMBER FUNCTION Unit 3 Unit 4 High-Head Safety Injection Check Valves 3-874A 4-874A Loop A, hot leg 4 3-875A. 4-875A- cold leg C 3-873A- 4-873A cold leg 3 3-874B 4-874B- Loop B, hot leg ,
cold leg-
~
3-875B 4-875B i 3-873B-- 4-873B cold leg ' 3-875C 4-875C Loop C', cold' leg 3-873C 4-873C cold leg .; Residual Heat Removal Line
. Check Valves
- 3-876-A 4-876A Loop A, cold leg 4-876E 3-8768 4-876B Loop B, cold leg 3-876D 4-876D 876C .4-876C' Loop'C, cold leg i 3-876E ,
4 M0V3-750 MOV4-750 Loop A, hot leg to RHR MOV3-751 MOV4-751 Loop A, hot leg to RHR
~
L ACCEPTABLE LEAKAGE LIMITS . 1
- 1. Leakage rates less than or equal to 1.0.gpm are considered acceptai>1e.
- 2. Leakage rates greater than 1.0 gpm but less tha'n or equal to 5.0 gpm are ;
considered acceptable provided that the latest measured rate has-.not exceeded the rate determined by the previous test by an amount.that reduces-the margin between previously measured leakage rate and the maximum permis-
- sible rate of 5.0 gpm by 50% or greater.
l 3.. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are \ l considered unacceptable if the latest measured-rate exceeded the rate l determined by the previous test by an amount that reduces the margin l , between previously measured leakage rate and the maximum permissible rate , ) of 5.0.gpm by 50% or greater. !
- 4. Leakage rates greater than 5.0 gpm are considered unacceptable.
TURKEY POINT - UNITS 3 & 4 3/4 4-22 AMENDMENT NOS.137AND 132 i t 9
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I. . 3/4.4.'7 ' CHEMISTRY l m 1 m
' JLIMITING CONDITION FOR OPERATION 8
1 g' . :q E3.'4;7/ The Reactor _ Coolant' System chemistry shall be maintained within the:
' limits specified in Table 3.4-2..
a
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- p ' APPLICABILITY: 'At all times.
.e ,
* -ACTION:
MODES 1,;2,,3 and 4:
.j
^ a. With any one or more chemistry parametec in excess of its Steady-c " : State Limit'but within its Transient Limit, restore the-parameter to within its, Steady-State Limit within 24 hours or be in at least HOT
- STANDBY within the next 6 hours and in COL 9 SHUTDOWN within the following 30' hours; and
- b. With any.one or more chemistry parameter in excess of its Transient 1
-Limit ~,'be in at least H0T. STANDBY within 6 hours and in COLD SHUTDCWN f within the following 30 hours.
1 At All Other Times:: l With the concentration of either chloride or fluoride in the Reactor j'< Coolant System in excess of its Steady-State Limit for more than 24 hours i
. or in excess of its Transient Limit, reduce the pressurizer pressure to l less than or equal to 500 psig, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the = ..-
. structural integrity of the Reactor Coolant System; determine that the H Reactor Coolant System-remains acceptable for continued operation prior - j to incre.asing the pressurizer pressure above'500 psig or prior to pro- 1
-ceeding to MODE 4.
1 I SURVEILLANCE REQUIREMENTS
-l i
L i4.4.7 The Reactor Coolant System chemistry shall be determined to be within
.the limits by analysis of those parameters at the frequencies specified in Table 4.4-3.
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1 t h L l- ~ .. l TURKEY POINT - UNITS 3 & 4 3/4 4-23 AMENDMENT NOS 137AND 132 l l~ 1
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V~ TABLE 3.4-2
'. - REACTOR COOLANT SYSTEM
.[ CHEMISTRY LIMITS STEADY STATE TRANSIENT PARAMETER LIMIT LIMIT
,:R r Dissolved Oxygen
- 1 0.10 ppm- .5 1.00 ppm Chloride **' 1 0.15 ppm 1 1.50 ppm; Fluoride ** 1 0.15 ppm ,-1 1.50 ppm z
e ,
"LimitnotaphF.licable equal to 250 with average reactor coolant temperature less than or
'**Not required when reactor is defueled and RCS forced circulation is
- unavai1able.
-TURKEY POINT - UNITS 3 & 4 3/4 4-24 AMENDMENT NOS.137 AND 132 f
i, I TABLE 4.4 ' REACTOR COOLANT SYSTEM: CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS. PARAMETER; SAMPLE AND ANALYSIS FREQUENCY Dissolved Oxygen
- At.least 5 times-per week not to.
exceed 72 hours between samples Chloride ** At least 5 times pe'r week not to exceed 72. hours between samples Fluoride **- At least 5 times per week not to exceed'72-hours between' samples F 0 i 7 1 . W. t "Not required with average reactor coolant temperature less than or equal to 250'F.
**Not required when reactor is defueled and RCS forced circulation is unavailable.
TURKEY POINT - UNITS 3 & 4 3/4 4-25 AMENDMENT N05137 AND 132
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REACTOR COOLANT SYSTEM-a ' 3/4.4.8 SPECIFIC ACTIVITY LIMITINC CONDITION FOR OPERATION 1 3.4.8 The~ specific-activity of the primary coolant shall be limited to:
- a. Less: than or equal to 1.0 microcurie per gram DOSE-EQUIVALENT I-131, and .
-b. Less than or equal to 1006 microcuries per gram of gross radioactivity. l APPLICABILITY: MODES 1, 2, 3, 4 and 5.
ACTION:- MODES 1, 2 and 3*: 3
- a. With the specific activity of the reactor coolant greater than 1
- , 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours- I L
during one continuous time interval or exceeding the limit line l L shown on Figure 3.4-1, be in at least HOT STAND 8Y with average I reactor coolant temperature less than 500*F within 6 hours; and j .: Lb. With the. specific activity of the reactor coolant greater than 100 R microcurie per gram, be in at least HOT STANDBY with average reactor coolant temperature'less than 500 F within 6 hours. , MODES 1,.2, 3, 4, and 5: With the specific activity of the reactor coolant greater than 1 micro-curie per gram DOSE EQUIVALENT I-131 or greater than 100M microcuries-per gram, perform the sampling and analysis requirements of Item 6.a) of l Table 4.4-4 until the specific activity of the reactor coolant is i l restored to within its limits. SURVEILLANCE REQUIREMENTS ; t 4.4.8 The specific activity of the reactor coolant shall be determined to be i within the limits by performance of the sampling and analysis program of y 1 Table 4.4-4. t i
*With the average reactor coolant temperature greater than or equal to 500 F.
TURKEY POINT - UNITS 3 & 4 3/4 4-26 AMENDMENT NOS137 AND 132 i
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i OPERATION \ - (80,60) l 50 i e - O 20 30 40 50 60 70 80 90 100 m RATED THERMAL POWER FIGURE 3.4-1 , DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram DOSE EQUIVALENT I-131. TURKEY POINT - UNITS 3 & 4 3/4 4-27 AMENDMENT N05.137 AND 132
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, R$ ACTOR COOLANT SPECIFIC ACTIVITY SAFPLE AND ANALYSIS PROGRAM [ l- ' l m - TYPE OF MEASUREMENT. -SAMPLE-AND ANALYSIS M00ES IN milch SAMPLE 4 h ANY ANALYSIS ' l- FREQUENCY AND ANALYSIS REQUIRED ,
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'- 1. Gross Radioactivity .At least..once per 1,2,3,4~ -
.g-Determination 72 hours.
- d 2. Tritium Activity 1 per 7 days. 1,2,3,41 -
w Determination o.
+ 3. Isotopic Analysis for 1 per 14 days. 1 l DOSE EQUIVALENT I-131 Concentration
- 4. Radiochemical Isotopic Monthly 1,2,3,4 g Determination Including Gaseous Activity l
E$ 5. Radiochemical for E 1 per 6 months" 1 Determination l l 6. Isotopic Analysis for a) Once per 4' hours, ' l#, 2#, 3#, 4#, 5# Iodine Including I-131, .whenever the I-133, and I-135 specific activity exceeds 1 pCi/ gram
> DOSE EQUIVALENT M I-131 or'100/E y pCi/ gram of gross g radioactivity, and w
z b) One sample between-1, 2, 3 8 2.and 6 hours; follow-' J t; ing~a THERMAL POWER change exceeding 15% E of the RATED-THERMAL i 5 . POWER within a 1-hour M , period.- -
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n'. 7n , D ;i L.m V, - Table ' 4 ~ 4-4 ' (Continued,), w y.s . - 1 ; t,, . ii A >- TdBLE NOTATIONS: , 9I
- Sample to be-taken;after,a minimum of.2 EFPD and'20 days of. POWER 0PERATION-have elapsed since reactor was last subcritical for 48 hours or longer.-
a f.:, : s
.#Until.the specific activity of the Reac' tor Coolant System is restored within' rf its limits.
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REACTOR COOLANT SYSTEM . 3/4 4.9 PRESSURE / TEMPERATURE LIMITS
/ REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature.and c pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 and 3.4-4 during heatup, cooldown, criticality, and inservice i leak and hydrostatic testing with:
- a. A maximum heatup of 100 F in any 1-hour period,
- b. A maximum cooldown of 100'F in any 1-hour period, and i
- c. A maximum temperature change of less than or equal to 5 F in any 1-hour period during inservice hydrostatic and leak testing opera- '
l tions above the-heatup and cooldown limit curves. APPLICABILITY: At all times. ACTIONi . With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-lim" condition on the structural integrity of the Reactor Coolant System; determina :qat the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the next 6' hours and reduce the RCS T,yg and pressure to less than 200 F'and 500 psig, respectively, within the following 30 hours. 4 SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall"be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determire changes in material properties, as required by 10 CFR Part 50, Appendix H in accordance with the sc'1edule in Table 4.4-5. The results of these exami#ations shall be used to update Figures 3.4-2, 3.4-3 and 3.4-4. . 4 i TURKEY POINT - UNITS 3 & 4 3/4 4-30 AMENDMENT N05.137 AND 132 l
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4 1 4 . M- MATERIAL- PROPERTY BASIS 1 00NTROLLING MATERIAL: CIRCUMfERENTIAL WELO l INITIAL RTNOT r 10*f
= SERVICE PERIOD: 20 EfPY RT e 1/4 THICKNESS = 252.5'f'-
NOT HEATUP RATES: UP TO 60*f/HR RT e 3/4 THICKNESS = 200.4'T NOT NOTE: NO HARGINS ARE GIVEN FOR POSSIBLE INSTRUMENT ERRORS. llllllllllll 1 l IEAK TEST LIMIT ll i !' tu l; l . , l, ' l l a 2000 ,
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l tr l a l 1 i 1M j l r r l E I CR l 1500 . / LIMIT D ON I INSERVI E HYDRO- i UNACCEPTABLE OPERATION j j STATIC TEST ; I1250 J
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. g 102 f ! PERIOD20.EFPY UP)TO 750
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f l HENLF WES UP TO / ! ACCEPTABLE::: p 500 60'f/HR
-M ' / l OPERATION ::: a 1
l ACCEPTABLE OPEPATION ! 2 l i 0 0 SO 100 150 200 250- 300 350 400 450- 500 l INDICATED TEMPERATURE (*F) l! i i FIGURE 3.4-2 TURKEY POINT UNITS 3 & 4 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS (60*F/hr) - APPLICABLE ' UP TO 20 EFPY TURKEY POINT - UNITS 3 & 4 3/4 4-31 AMENDMENT N05.137 AND 132
c .; MATERIAL PROPERTY BASIS j l CONTROLLING MATERIAL: ORCUMFERENTl/l WELD i itolTLAL RTNDT: 10*f j SERVICE PLtl0D: 20 EFPY RT NDT e 1/4 THICKNESS = 252.5'F HEATUP RATES: UP TO 100*F/HR RT e 3/4 THICKNESS = 200.4*F NDT NOTE: NO MARGINS ARE OfVEN FOR POSSIBLE INSTRUMENT ERRORS.
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2" ' ' H !!!lllllllll , LEAK TEST LIMIT T f - - F---- ; 22 % ; j j ,
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2000 ]
/i CRmCAUTY 1E l f; LIMIT (BASED ON j j' lNSERVICE HYDRO-r r STATIC TEST ,
l i TEMPERATURE L UNACCEPTABLE OPERATION I} R SERVC( 1250 l l PERIOD UP TO , j '--~20 EFPY e I
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ACCEPTABLE:
. / OPERATION -
- HEATUP R TES --- -
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UP TO 100'F/HR'
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500 llll ll s 250 ACCEPTABLE [ OPERATION g I O 50 100 150 200 250 300 550 400 450 500 INDICATED TEMPERATURE (*F) , l i l FIGURE 3.4-3 TURKEY POINT UNITS 3 & 4 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS (100*F/hr) - APPLICABLE UP TO 20 EFPY TURKEY POINT - UNITS 3 & 4 3/4 4-32 AMENDMENT N05.137 AND 132
f, MATERIAL PROPERTY BASIS CONTR0gING MA]' RIAL: CIRCUMFERENTIAL WELO T NTIAL R NDT: 1 F NDT e1/4 THICKNESS = 252.5T s SERVICE PERIOD: 20 EFPY RT e3/4 THICKNESS = 200.4T NDT COOLDOWN RATES: UP TO 100*F/HR NOTE: NO MARGINS ARE GIVEN FOR POSSIBLE INSTRUMENT ERRORS. 2500 i . r
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2250 $ I 2= L_. l A
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1500 1 UNACCEPTABLE / OPERATION f 1250 / r j --- som / r COOLDOWN e -
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O 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERAT .;E (*F) FIGURE 3.4-4 TURKEY POINT UNITS 3 & A REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS (100'F/hr) - APPLICABLE UP TO 20 EFPY TURKEY POINT - UNITS 3 & 4 3/4 4-33 AMENDMENT N05.137 AND 132
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TABLE 4.4$is -i , h ' REACTOR' VESSEL MATERIAL SURVEILLANCE PRJGRAM - WITilDRAWAL SCHEDULE :l UNIT 3 2; . . 1 CAPSULE- VESSEL LEAD- ,
, NUMBER LOCATION FACTOR .n'ITHDRAWAL TIME - ,
(( U: F. ' O.49_ ..itandby l,
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,:; at'12 years :l 2 . ,, W- 40' O 34 -Standby g-- j g X 50 - 0.34 33' years ) Y 150' O.49 Standby ,!, Z' 230' O.34 Standby UNIT 4 j L,; CAPSULE- VESSEL LEAD f WITHDRAWAL TIME l NUMBER, LOCATION FACTOR U 30 0.49 Standby 1 V 290' O.79 '24 years, , j W- 40': 0.34 SLndby-. j
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!-X 50 ' O.34 Standby ,_f Y 150 0.49 Standby- .
Z 230 0.34 Standby 1 1 pa, ,j J e s n r - TURKEY POINT'- UNI 15 3 & 4 3/4 4-34 AMENDMENT NOS237 AND132 , t
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' PRESSURIZER. :
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. LIMITING CONDITION FOR OPERATION p
v j '3.4.9.2 The pressurizer temperature shall be limited to: 0. j a. A maximum heatup of 100 F in any 1-hour period, d b .- .A maximum cool.down of 200 F in any 1-hour period, and {. b , c. A maximum spray water temperature differential of 320'F. ' i
' APPLICABILITY: At all. times.
_ ACTION: With the pressurizer temperature limits in excess of any of the above limits, I restore.the temperature to within the limits within 30. minutes;. perform en engineering, evaluation to determine the effects of the out-of-limit condition on the structural integrity-of the pressurizer; determine that the pressurizef rema. ins' acceptable for continued operation or be in at least HOT. STANDBY within the next 6 hours a.id reduce the pressurizer pressure to less than 500 psig with-in the following 30 hours.
~
t juRVEILLANCEREQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the , limits at least once per 30 minutes during system heatup or cooldown. lne spray water temperature'dif ferential shall be determined to be within the H mit at least once per 12 hours during auxiliary spray operation. .j 4 i-s 1 l TURKEY POINT - UNITS 3 & 4 3/4 4-35 AMENDMENT N05.137AND 132 4
I. ' pa > l L REACTOR COOLANT SYSTEM iT n. OVERPRESSURE MITIGATING SYSTEMS . l LIMITING CONDITION FOR OPERATION [.- 3.4.9.3 The_highpressuresafetyinjectionflowpathstotheReactorCoolant . System (RCS) shall be isolated, and below an RCS average coolant temperature ! of 275'F at least one of the following Overpressure Mitigating Systems shall be OPERABLE: b.*
'a) Two power-operated relief valves (PORVs) with a lift setting of ,
415 1 15 psig, or
~
i b) The RCS depressurized with a RCS vent of greater than or equal to- , 2.20 square inches, j APPLICABILQY: MODES 4, 5 and 6 with the reactor vessel head on. ACTION:
- a. Withthe-highpressuresafetyinjectionflowpathstotheRCS !
unisolated, restore isolation of these flow paths within 4 hours, and
- b. In MODE 4 with RCS average coolant temperature less than or ecual to 275 F, and in MODE 5 or in Mode 6 with the reactor vessel heac on:
- 1. Wit'i one PORV. inoperable, perform at least one of the following ;
wi' Sin the next 7 days:- ' L a) Restore the inoperable PORV to OPERABLE status, or [ t b) Depressurize and vent the RCS through at-least a 2.20 square inch vent, or c) Depressurize and maintain a RCS: vent'through at least one-open PORV and open associated block valve.
- 2. With both PORVs inoperable, depressurize and vent the RCS .
through at least a 2.20 square inc.h vent within 24 hours.
- 3. In the event either the PORVs or a 2.20 square inch vent-is used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specifica-tion 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence, t-h L - TURKE.Y. POINT - UNITS 3 & 4 3/4 4-36 AMENDMENT NOS.137AND ^.'
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i' REACTOR COOLANT SYSTEM i DVERPRESSURE MITIGATING SYSTEMS SURVEILLANCE REQUIREMENTS __ 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:
- a. Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actua-tion channel, but excluding valve operation, within 31 days prior to '
entering a condition in which the PORV is required OPERABLE and at least once per'31 days thereafter when the PORV is required OPERABLE.
! b. Performance of a CHANNEL CAllBRATION on the PORV actuation channel at least once per 18 months; and
- c. Verifying the PORV block valve is open at least once per 72 hours when the PORV is being used for overpressure protection,
- d. While the PORVs are required to be OPERABLE, the backup air supply shall be verified OPERABLE at least once per 24 hours, 4.4.9.3.2 The 2.20 square inch vent shall be verified to be open at least once per 12 hours
- when the vent (s) is being used for overpressure i protection.
- 4.4.9.3.3 Verify the high pressure injection flow path to the RCS is isolated
.at least once per 24 hours by closed valves with power removed or by locked 1
- closed manual valves, q l
I t 1
*Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.
i TURKEY POINT - UNITS 3 & 4 3/4 4-37 AMENDMENT NOS.137 AND 132 l l
9, l REACTOR COOL M ' SYSTEM g 3/4.4.10 STRUCTURAL INTEGRITY t i I LIMITING CONDITION FOR OPERATION L 3.4.10 The structural integrity of ASME Code Class 1, 2 and 3 components shall l L be maintained in accordance with Specification 4.4.10. i>
~
L All MODES APPLICABILITY: i ACTION:- -
- a. With the structural integrity of any ASME Code Class I component (s)-
not conforming to the above requirements, restore the structural
- e integrity of the affected component (s) to within its limit or isolate ,
the affected component (s) prior to increasing the Reactor Coolant ! System temperature scre than 50 F above the minimum temperature- ! required by NDT considerations,
- b. With~the structural integrity of any ASME Code .' lass,2 component (s) j not conforming to the above requirements, restsre the structural
^
integrity of the affected component (s) to within its limit or isolate i the affectet >mponent(s) prior to increasing he Reactor Coolant System temperature above 200 F. l
- c. With the structural integrity of any ASME Code Class 3 component (s) '
not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate- ' the affected component (s) from service. i SURVEILLANCE REQUIREMENTS ; 4.4.1'O ~ 1n addition to the requirements of Specification 4.0.5, each reactor , coolant pump flywheel shall be inspected per the recommendations of Regulatory Position _C 4,b of Regulatory Guide 1.14, Revision 1, August 1975. TURKEY POINT - UNITS 3 & 4 3/4 4-38 AMENDMENT N05.137AND 132 l
~
1 k REACTOR COOLANT SYSTEM 3/4.4.11 REACTOR COOLANT SYSTEM VENTS l LIMITING CONDITION FOR OPERA't0N
- 3.4.11 At least one Reactor Coo nnt System vent path consisting of at least '
two vent valves in series and powered from emergency busses shall be OPERABLE i and closed at each of the following locations:
- a. Reactor vessel head, and
- b. Pressurizer steam space j APPLICABILITY:
MODFS 1, 2, 3, and 4. . ACTION: ,
- a. With one of-the above Reactor Coolant system vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable-vent path is maintained closed with power removed from the valve .
actuator of all the vent valves in the inoperable vent path; restore , the inoperable vent path to OPERABLE status within 30 days, or, be i in HOT STANDBY within 6 hours and in COLD 5HUTDOWN within the follow- ; ing 30 hours,
- b. With both Reactor Coolant System vent paths inoperable; maintain the :
inoperable vent path closed with power removed from the. valve actua- t tors of all the vent valves in the inoperable vent paths, and restore l at least one of the vent paths to OPERABLE status within 72 hours or be in HOT STANDBY within 6 hours and in COLD SHUfDOWN within the following 30 hours. CURVEILLANCE REQUIREMENTS 4.4.11 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at , least once per 18 months by: ;
- a. Verifying all manual isolation valves in each vent path are locked in the open position, '
- b. Cycling each vent valve through at least or'e complete cycle of full travel from the control room, and
- c. Verifying flow through the Reactor Coolant System vent paths during venting, TURKEY POINT - UNITS 3 & 4 3/4 4-39 AMENDMENT NOS137 AND 13?
(
i 1 3/4.5 EMERGEMOY CORE COOLING SYSTEMS 1 3/4. 5.1 ACCUMULMUn*, LIMITING CONDITION FOR OPERATION , 3.5.1 Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:
- a. The isolation valve open and its circuit breaker open, ,
- b. An indicated borated water volume of between 6545 and 6665 gallons,
- c. A boron concentration of between 1950 and 2350 ppm, i
- d. A nitrogen cover pressure of between 600 and 675 psig, and
- e. A water level and pressure channel OPERABLE.
1 1 l APPLICABILITY: MODES 1, 2, and 3*. l. ACTION: l
- a. With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 4 hours or be in at least HOT STANDBY within the next 6 hours and reduce pressurizer pressure to less than 1000 psig within the followinD 6 hours.
l b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolatior, valve or be in at least HOT STAN09Y within 6 hours and reduce pressurizer pressure to- ' less than 1000 psig within the following 6 hours. 1 SURVEILLANCE REQUIREMENTS , l-4.5.1.1 Each accumulator shall be demonstrated OPERABLE:
- a. At least once per 12 hours by:
l l
- 1) Verifying the indicated borated water volume and nitrogen ,
cover pressure in the tanks, and
- 2) Verifying that each accumulator isolation valve is open by control room indication (power may be restored to the valve l operator to perform this surveillance if redundant indicator is
- inoperable).
- Pressurizer pressure above 1000 t.sig.
TURKEY POINT - UNITS 3 & 4 3/4 5-1 AMENDMENT NOS.137AND 132
f~ ,
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[. EMERGENCY CORE COOLING SYSTEMS -j F l L! "
, SURVEILLANCE REQUIREMENTS-(Continued)
(* ! i b. At least once per 31 days and within,6 hours after each sclution i volume increase of greater than or equal to 1% of tank ,11ume by l l- verifyi g the boron concentration of the solution in the ' F water-f 11ed' accumulator; l .
- c. At least.once per 31 days: .
j
- 1) When the RCS pressure is above_1000 psig,-by verifying that the
!' power to the isolation valve operator is disconnected by a j locked open breaker, and i
. t t> -
- 2) Each accumulator water _ level and pressure channel shall be demonstrated OPERABLE by the performance of an ANALOG CHANNEL 1 OPERATIONAL TEST, and ;
- e d.- At least once per 18 months: ,
- 1) Each accumulator water level and pressure channel shall be ,
demonstrated OPERABLE by the performance of a CHANNEL , CAllBRATION,.and- :
- 2) Each accumulator check valve:shall be checked for operability. ] ;
J I i i
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k 4 TURKEY POINT - UNITS 3 & 4' 3/4 5-2 . AMENDMENT NOS237 AND 132 1
EMERGENCY CORE COOLING SYSTEMS 3/4.5.2 ECCS SUBSYSTEMS - T,yg GREATER THAN OR EQUAL TO 350*F LIMITING CONDITION FOR OPERATION 3.5.2 The following Emergency Core Cooling System (ECCS) equipment and flow paths shall be OPERABLE:
- a. Four OPERABLE Safety injection (SI) pumps with discharge aligned to the RCS cold. legs,
- b. Two OPERABLE RHR heat exchangers,
- c. Two OPERABLE RHR pumps with discharge aligned to the RCS cold legs,
- d. An OPERABLE flow path capable of taking suction f rom the refueling water stcrage tank as defined in Specification 3.5.4, and
- e. Two OPERABLE flow paths capable of tak'ng suction from the containment sump.
ADPLICABILITY:' MODES 1, 2, and 3*. ACTION:
- a. With any one of the required ECCS components or flow paths inoperable, except for inoperable Safety injection Pump (s), restore the inoperable component or flow path to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours,
- b. In the event the ECCS is actuated and injects water in the ne utor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date since January 1, 1990.
- c. With one Safety Injection pump inoperable, restore the pump to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 12 hours and in HOT SHUTDOWN within the following 6 hours.
This ACTION applies to both units simultaneously,
- d. With two Safety Injection Pumps inoperable, restore one of the two inoperable pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 12 hours and in HOT SHUTDOWN within the following 6 hours. This ACTION applies to both units simultaneously.
*The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry into MODE 3 for the safety Injection flow paths isolated pursuant to Specifica-tion 3.4.9.3 provided that the Safety injection flow paths are restored to OPERABLE status prior to T exceeding 380 F. Safety Injection flow paths may avg be isolated when T avg is less than 380 F.
TURKEY POINT - UNITS 3 & 4 3/4 5-3 - AMENDMENT NOS137 AND 132
a;; M ay-EMERGr.NCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each.ECCS component and flow path shall be demonstrated OPERABLE:
- a. At least once per 12 hours by verifying by control room indication that the following valves are in the indicated positions with power to the valve operators removed: '
Valve Number Valve Function Valve Position 864A and B Supply from RWST to ECCS Open 862A and B RWST Supply to RHR pumps Open
. 863A and B RHR Recirculation Closed 866A and B H.H.S.I. to Hot Legs Closed HCV-758* RHR HX Outlet Open, To permit temporary operation of these velves for surveillance or .
maintenance purposes, power may be restored to these valves for a i period not to exceed 24 hours,
- b. At least once per.31 days by:
- 1) Verifying that the ECCS piping is full of water by venting the ECCS pump casings and accessible discharge piping,
- 2) Verifying that each valve (manual, power-operated, o* automatic) in the flow path that is not locked, sealed, or ctherwise secured in position, is in its correct position, and !
- 3) Verifying that each RR'l Pump develops the indicated differential pressure applicable to the operating conditions in accordance with Figu.e 3.5-1 when tested pursuan to Specification 4.0.5.
c, At least once per 92 days by: l 1) Verifying that w h SI pump develops the indicated differential pressure appli sie to the operating conditions when tested j pursuant to specification 4.0.5. SI pump > 1126 psid at a metered flowrate > 300 gpm (ncrmal' alignment or Unit 4 SI pumps aligned to Unit 3 RWST), or
> 1156 psid at a metered flowrate > 280 gpm IUnit 3 SI pumps aligned to Unit 4 RFST).
l 1
- Air Supply to HCV-758 shall be verified shut off and sealed closed once per 31 days. 1 TURKEY POINT - UNITS 3 & 4 3/4 5-4 AMENDMENT N05.137 AND 132
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5 y , t 0 1000 2000 3000 400C RHR PUMP FLOW (GPM) 1 t Figure 3.5-1 RHR Pump Curve TURKEY POINT - UNITS 3 & 4 3/4 5-5 AMENDMENT NOS.137AND 132
m E G p.' I EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQti!REMENTS
- d. By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:
- 1) For all acc sible areas of the containment prior to establish-ing CONTAINhulT INTEGRITY, and L 2) Of the areas affected within containment at the completion of each containment entry when CONTAINMENT;1NTEGRITY is
' established.
- e. At least once per 18 months by:
- 1) Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System by ensuring that with a simulated or actual Reactor Coolant System pressure signal.
greater than or equal to 525 psig the interlocks cause the
# valves to automatically close and prevent the valves from being opened, and
- 2) Verifying correct interlock action to ensure that'the RWST is isolated from the RHR System during RHR System operation and to ,
ensure that the RHR System cannot be pressurized from the ! Reactor Coolant System unless the above RWST Isolation Valves l are closed. ; 3)- A visual inspection of the containment sump and verifying that the suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.
- f. At least once per 18 months, during shutdown, by:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position on Safety injection actuation test signal, and
- 2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection actuation test signal:
a) Safety Injection pump, and b) RHR pump. TURKEY POINT - UNITS 3 & 4 3/4 5-6 AMENDMENT NOS.137AND 132
l EMERGENCY CORE COOLING SYSTEMS' SURVEILLANCE REQUIREMENTS ,,
- g. By verifying the correct position of each electrical and/or mechanical position stop fo the following ECCS throttle valves:
- 1) Within 4 hours followir.. completion of each valve streking operation or maintenance on the valve when the ECOS components are required to be OPERABLE, and
- 2) At least once per 18 months.
RHR System Valve Number , HCV *-758 MOV *-872 FCV *-605 s 3/4 5-7 AMENDMENT N05.137AND 132
- TURKEY POINT - UNITS 3 & 4
l EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS T,yg LESS THAN 350'F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, the following EC's components and flow path shall be OPERABLE: s
- a. One OPERABLE RHR heat exchanger,
- b. One OPERABLE RHR pump, and
- c. An OPERABLE flow path capable of (1) taking suction from the refueling water storage tank upon being manually realigned and (2) transferring suction to the containment sump during the recirculation phase of opera' tion.
1 APPLICABILITY: MODE 4. ACTION:
- a. With no OPERABLE ECCS flow path from the refueling water storage .n tank, restore at least one ECCS flow path to OPERABLE status within I hour or be in COLD SHUTOOWN within the next 24 hours,
- b. With either the residual heat removal heat exchanger or RHR pump I inoperable, restore the components to OPERABLE status or maintain the Reactor Coolant System T avg less then 350 F by use of alternate !
heat removal methods, l c. In the event the ECCS is actuated and injects water into the Reactor Coolant Syster a Special Report.shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date since January 1, 1990. SURVEILLANCE REQUIREMENTS 4.5.3 The above ECCS components shall be demonstrated OPERABLE per the
. applicable requirements of Specification 4.5.2. !
)
3 i h- i TURKEY POINT - UNITS 3 & 4 3/4 5-8 AMENDMENT N05.137 AND 132 i I A
EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.4 For single linit operation, one refueling water storge tank (RWST) shall be OPERABLE or for dual Unit operation two RWSTs shall be OPERABLE with:
- a. A minimum indicated borated water volume of 320,000 gallons per RWST,
- b. A minimum boron concentration of 1950 rpm of boron,
- c. A minimum solution temperature of 39'F, and
- d. A maximum solution temperature of 100 F.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With less than the requ, red num'ar of RWST(s) OPERABLE, restore the tank (s) to OPERABLE status within i hour on be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.5.4 The required RWST(s) shall t. Jaaenstrated OPERAE;1:
- a. At least once per 7 days by:
- 1) Verifying the indicated borated water volume in the tank, and
- 2) Verifying the boron concentration of the water,
- b. By verifying the RWST temperature is within limits whenever the outside air temperature is less than 39 F or greater than 100 F at the following frequencies:
- 1) Within one hour upon the outside temperature exceeding its limit for consecutive 23 hours, and I
- 2) At least once per 24 hours while the outside temperature '
exceeds its limit. i 1 TURKEY POINT - UNITS 3 & 4 3/4 5-9 AMENDMENT N05.137 AND132 l
'a- ,
1 ai
'3/4.6 CONTAINMENT SYSTEMS L
3/4.6.1 PRIMARY CONTAINMENT ; CONTAINMENT INTEGR1TY LIMITING CONDITION FOR OPERATION j i l 3.6.1.1 Primary CONTAINMENT INTEGRITY shall be maintained * ! APPLICABILITY: MODES 1, 2, 3, and 4. l I ACTION: l Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within
'I hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.6.1.1 CONTAINMENT INTEGRITY shall be demonstrated:
- a. At least once per 31 days by verifying that all penetrations ** not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident cc,nditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions;
- b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and L
- c. After each closing of each penetration subject to Type B testing, except the containment air locks, if opened following a Type A or B-test, by leak rate testing t.he seal with gas at a pressure not less than 50 psig, and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specification 4.6.1.2d. for all other Type B and C penetrations, -
the combined leakage rate is less than 0.60 L,,
- Exception may be taken under Administrative Controls for opening of valves and airlocks necessary to perform surveillance, testing requirements and/or corrective maintenance. In addition, Specification 3.6.4 shall be complied with.
**Except valves, blir.d flanges, and deactivated automatic valves which are ;
located inside the cantainment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need not be performed more often than once per 92 days, i TURKEY POINT - UNITS 3 & 4 3/4 6-1 AMENDMENT N05.137AND 132
.o
o s CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE-
- c. LIMITING CONDITION FOR OPERATION p 3. 6.1. 2 Containment leakage rates shall be limited to:
- a. An overall integrated leakage rate of less than or equal to L p ,
~
0.25% by weight of the containment air per 24 hours at P,, 49.9 psig. p b. A combined leakage rate of less than 0.60 L, for all penetrations { andvalvessubjecttoTypeBandCtests,whenpressurizedto ' P,. APPLICABILITY:. MODES 1, 2, 3, cnd 4. ACTION: With either the measured overall integrated containment leakage-rate exceeding . i-
~'
0.75 L, or the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 t,, restore the overall ; integrated leakage: rate to less than 0.75 L, and the combined leakag0 rate for' ' all- penetrations subject to Type B and C tests to less than 0.60 L, prior to , increesing the Reactor Coolant System temperature above 200"F. i SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the followilg test schedule and shall be determined in conformance with the criteria speci- , fied in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI N45.4-1972:=
- a. Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40110 month intervals during shutdown at'. '
a pressure not less than P,, 49.9 psig, during each 10-year service period. The third test of each set shall be conducted during the I shutdown for the 10 year plant inservice inspection; !
+
t
!TURKEYPOINT_--UN1153&4 3/4 6-2 AMENDMENT N05.137 AND 13?- g
l l CONTAINMENT SYSTEMS SURVEILLANCEREQOIREMENTS(Continued) '
- b. If any periodic Type A test fails to meet 0.75 L, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission. If two consecutive Type A tests fail to meet 0.75 L,,
l a Type A test shall be performed at least every 18 months until two b consecutive Type A tests meet 0.75 L, at which time the above test i schedule may be resumed; 1 c .- The accuracy of each. Type A test shall be verified by a supplemental test which:
- 1) Confirms the accuracy of the test by verifying that the supple- ,
mental test result, L c, is in accordance with the following ' equation: lL c- (L,, + L )gl 5 0. 25 L, where L am is the measured Type A test leakage ando t is the , superimposed leak; '
- 2) Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; ,
, and
- 3) Requires that the rate at which gas is injected into the contain-ment or bled from the containment durlng the supplemental test be limited to between 0.75 La and 1.25 L,;
- d. Type 3 and C tests shall be conducted with gas at a pressure not less than F,, 49.9 psig, at intervals no greater thaa 24 months except fu Wats involving:
- 1) Air locks,
- 2) Purge supply and exhaust isolation valves, and
- 3) Equipment access opening which shall be tested at least once every 12 months and after each use,
- e. Air locks shall be tested and demonstrated OPERABLE by the require-ments of Specification 4.6.1.3;
- f. Purge supply and exhaust isolation valves seals shall be tested and demonstrated OPERABLE by the requirements of Specification 4.6.1.7.2, as applicable;
- g. The provisions of Specification a.0.2 are not applicable.
l TURKEY POINI - UNITS 3 & 4 3/4 6-3 AMENDMENT N05.137 AND 132
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. CONTAINMENT SYSTEMS l c- '
, CONTAINMENT AIR LOCKS 4 c
]
i g LIMITING CONDITION F0,R OPERATION I g, - 3. 6.'1. 3 Each containment air lock shall be OPERABLE with: p, a. Both doors closed except when the air lock is being used for. normal
, . transit entry and exit through the containment, or during the .
performance of containment air lock surveillance and/or testing J L requirements; then at least one air lock' door shall be closed, and b 'b. An overall air lock leakage rate of less than or equal to 0.05 L, at P,, 49.9 psig. p' APPLICABILITY:' MODES.1, 2, 3, and 4. ; ACTION: o ' a.- With one containment air lock door inoperable:
- 1. - ' Maintain at least the OPERABLE air lock door closed and either !
restore the inoperable air lock door to OPERABLE status within i 24 hours or lock the OPERABLE air lock door closed;. 1
- 2. Operation may then continue until performance of'the next .!
required overall air lock leakage test provided that'the , OPERABLE air lock door is verified to be locked closed at a least once per 31 days;
- 3. 0therwise, be in at least HOT STANDBY within'the next 6 hours '
and in COLD SHUTDOWN within the following 30 hours.
~
With the containment air lock inoperable' except as the result of ant
~
- b. !
inoperable air lock door, maintain at least one air lock door closed;
. restore the inoperable air lock to OPERABLE status within 24 hours !
or be in at least HOT STANDBY within the next 6 hours and in COLO' SHUTDOWN within the following 30 hours. F
'l.
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'T-TURKEY. POINT -- UNITS 3 & 4 3/4 6-4 AMENDMENT NOS,137At:') 132 .l J ' ([,
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9"7F'7pE ,"(' n c .. 1 si , l CONTAINMENT SYSTEMS . j i
- SURVEILLANCE REQUIREMENTS' l
c 4. 6.1. 3 Each coi.: enment air lock shall be demonstrated OPERABLE:- ] i a.. .Within 72 hours following each closing, except when the air lock is -l being used for multiple entries, then at least once per 72 hours, by verifying that the seals have not been damaged >and have seated 4
- properly by. vacuum testing the volume between the door seals in- '
- l
+ accordance with approved plant procedures. ci q
- b. By conducting overall air lock. leakage tests at notiless 'than' i 50 psig. .and verifying the overall air lock leakage rate is within '
its limit at.least once per 6 months.*- -j lc. At least once per 6 months byEverifying that only one door in each
- c ' air lock can be' opened at a time.
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*The_p'rovisions of-Specification 4.0.2 are not' applicable, TURKEY POINT - UNITS 3 & 4 3/4.6-5 AMENDMENT 1405.137AND 132 u
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s n 1 CONTAINMENT SYSTEMS l - , 1 ' INTERNAL ~ PRESSURE' l LIMITING' CONDITION FOR' OPERATION ,,_ l 3.6.1.4 Primary containment internal pressure shall be maintained between and +3.psig. 9 APPLICABILITY: MODES 1, 2, 3, and 4. l2 ACTION: , With the containment' internal pressure outside of the limits above, restore $ the internal pressure to within the' limits within 1 hour or be in atfleast HOT ; STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 Fours.- , ,. l r k
- s SURVEILLANCE REQUIREMENTS ls y
s 4.6.1.4. The primary containment internal pressure shall be determihed to be within the limits at least once per 12 ho'urs. 7 f f T'i i ( E : p' TURKEY POINT UNITS 3 & 4 3/4 6-6 AMENDMENT N05.137AND 132 b k l . - , ,, . ,,
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.g' CONTAINMENT SYSTEMS I
AIR: TEMPERATURE ,
-5
'l L , LIMITING CONDITION FOR OPERATION .
p . l' ~ 3. 6.1. 5 Primary containment' average air temperature shall not exceed 125'F
- and shall not exceed 120*f by more than 336 equivalent hours
- during a calendar year. ;
APPLICABILITY: MODES 1, 2, 3, and 4. :
' ACTION:
n a .' ' 1
'With the containment average air tempetature greater than 125 F or greater than.
' 120'f for more than' 336 equivalent hours
- during a calendar year, reduce.. the. ,,
or be f F average air. temperature to within the applicable limit within'8 hours,ithin in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN w L-(? the following 30 hours. r L U , SURVEILLANCE REQUIREMENTS-L o
~
4.6.1.5 The primary containment average air temperature shall be the arith -- metical ~ average of the temperatures at the following locations and.shall be : determined at least once per 24 hours: , Approximate Location
- a. 'O Azimuth 58 feet elevation .
- b. 120 Azimuth- 58 feet elevation- -;
I
- c. 240 Azimuth 58 feet elevation i .
- Equivalent hours are determined from actual hours using the time-temperature relationships that support the environmental qualification requirements of
.10 CFR'50.49.
g l TURKEY POINT - UNITS 3 & 4 3/4 6"7 AMENDMENT N05.137AND 332 l 1 e
r- . CONTAINMENT SYSTEMS l CON 1AINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment shall be maintained at a l level consistent with the acceptance criteria !n Specification 4.6.1.6. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With more than one tendon (not including exempted
- tendons) with an observed lift-off force between the predicted lower limit and 90%
of the predicted lower limit or with one tendon below 90% of the predicted lower limit, restore the tendon (s) to the required level of integrity within 15 days and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 30 days in accordance with Specification 6.9.2 or be in at - least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within T the following 30 hours. -
- b. With one exempted
- tendon with an observed lift-off force at the accessible end below 86% of the predicted lower limit, restore the tendon to the required level of integrity within 15 days and perform '
an engineering evaluation of the containment and provide a Special Report to the Commission within 30 days in accordance with Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
- c. With any abnormal degradation of the structural integrity other than ACTION a at a level below the acceptance criteria of Specifications 4.6.1.6.1, 4.6.1.6.2 and 4.6.1.6.3, restore the containment to the required level of integrity within 72 hours and perform an engineering evaluation of the containment and provide a Special Report to the Commission within 15 days in accordance with Specification 6.9.2 or be in at least HOT STANDBY within the next 6 hourt and in COLD ,
SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 1 .: 4.6.1.6.1 Containment Tendons. The containment tendons' structural integrity shall be demonstrated every fif th year from the date of the initial structural l' integrity test. The tendons' structural integrity shall be demonstrated by:
^ Exempted in accordance with IWL-2521.1(a). Lift-off forces observed at the
- accessible end below 90% of the predicted lower limit shall be reported to the Commission for information only.
TURKEY POINT - UNITS 3 & 4 3/4 6-8 AMENDMENT N05.137AND 132 xit
( i i CONTAINMENT SYSTEMS i SURVEILLANCE REOUIREMENTS (Continued)
- a. Determining that a random but representative sample ** of at least 12 tendons (3 dome, 4 vertical, and 5 hoop) each have an observed lift-off .
i force within predicted limits for each. For each subsequent inspection one tendon from each group may be kept unchanged to develop a history and.to correlate the observed data. If the observed lift-off force inal ' of any one sample tendon (not population lies including between exempted
- tendons) the predicted in the lower limit andorig 90% of the predicted lower limit, two tendons, one on each side of this tendon
!( , should be checked for their lift-off forces. Ifbothoftheseadjacent tendons are found to be within their predicted limits, all three tendons should be restored +o t.he required level of integrity. This single deficiency rtay P > W ed unique and acceptable. l' n
! b. Per arming te ' " -
'ng, inspections, and material tests on a previously ( , ' om each group (dome, vertical, and hnop).
A randomly se. - a m each group shall be completely detensi ened in order to id, . . <r damaged wires and determining that over the entire >
~'
a removed wire or strand that:
- 1) The tendon wires or strands are free of corrosion, cracks, and damage,
- 2) There are no changes in the presence or physical appearance of the sheathing filler grease, and
- 3) A minimum tensile strength of 240,000 psi (guaranteed u'ltimate strength of the tendon meterial) for at least three wire or strand samples (one from each end and one at mid-length) cut l
"' from each removed wire or strand. Failure of any one of the wire or strand samples to meet the minimum tensile strength test is evidence of abnormal degradation of the containment structure. 4
- c. Performing tendon retensioning of tho;e tendons detensioned for inspection to their observed lif t-off force with a tolerance limit o f + 6, -0%. During retensioning of these tendons, the changes in load and elongation should be measured simultaneously at a minimum of three approximately equally spaced levels of force between zero and the seating force. If the elongation corresponding to a specific '
load differs by more than 5% from that recorded during installation, an investigation should be made to ensure that the difference is not . related to wire failures or slip of wires in anchorages;
- Exempted in accordance with IWL-2521.1(a). Lift-off forces observed at the accessible end below 90% of the predicted lower limit shall be reported to the l Commission for information only. l L **After the process of randomly selecting tendons is performed, inaccessible tendons may be exempted in accordance with IWL-2521.1(a). Substitute tendons shall be selected that are located as close as possible to the exempted tendons.
The accessible end of exempted tendons shall have the lif t-of f force measured. TURKEY POINT - UNITS 3 & 4 3/4 6-9 AMENDMENT N05.137AND 132
F CONTAINMENT SYSTEMS SURVEILLANCE RE001REMENTS (Continued)
- d. Assuring that the observed lif t-of f force for each tendon exceeds the minimum required lift-off force. Required lift-off forces shall be calculated individually for each surveillance tendon prior to the beginning of each surveillance, and should consider such factors as:
- 1) Prestressing history;
. 2) Friction losses; and .
- 3) Time-dependent losses (creep, shrinkage, relaxation), considering time elapsed from prestressing.
- e. Verifying the OPERABILITY of the sheathing filler grease by:
- 1) Minimum grease coverage exists for the different parts of.the anchorage system, and
- 2) The chemical properties of the fii;er material are within the tolerance limits as specified by the manufacturer.
4.6.1.6.2 End Anchorages and Adjacent Concrete Surfaces. The structural integrity of the end anchorages of all tendons inspected pursuant to Specifi-cation 4.6.1.6.1 and the adjacent concrete surfaces shall be demonstrated by determining through visual inspection that no unacceptable levels of corrosion exist on the end anchorages and no unacceptable cracking exists in tha concrete adjacent to the end anchorages. Determination of acceptance levels shall be by engineering evaluation of the areas in question. If unacceptable conditions are found, the tendons inspected during the previous surveillance shall be examined to determine whether the corrosion levels or concrete cracking have increased since the previous surveillance. Inspection of adjacent concrete surfaces shall be performed concurrently with the containment tendon surveillance (Technical Specification 4.6.1.6.1). 4.6.1.6.3 Containment Surfaces. In accordance with 10 CFR 50, Appendix J. Section V. A, a visual inspection of the accessible interior and exterior surfaces of the containment, including the liner plate, shall be performed during the shutdown for (but prior to) each Type A containment leakage rate test (Technical Specification 4.6.1.2.1). The purpose of this inspection shall be to identify any evidence of structural deterioration which may ' affect containment structural integrity or leaktightness. The visual inspection shall be general in nature; its intent shall be to detect gross areas of widespread cracking, spalling, gouging, rust, weld degradation; or grease leakage. The visual examination may include the utilization of binoculars & other optical devices. Corrective actions taken, and recording of. structural deterioration and corrective actions, shall be in accordance with 10 CFR 50, Appendix J, Section V. A. Records of previous inspections shall be reviewed to verify no apparent changes in appearance. The first
,nspection performed will form the baseline for future surveillances.
TURKEY P0lNT - UNITS 3 & 4 3/4 6-10 AMENDMENT N05.137AND 132
L CONTAINMENT SYSTEMS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.6.1.7 Each containment purge supply and exhaust isolation valve shall be OPERABLE and:
- a. The containment purge supply and exhaust isolation valves shall be sealed closed to the maximem extent practicable but may be open for
-purge system operation for pressure control, for environmental conditions control, for ALARA and respirable cir quality considera-tions for personnel entry and for surveillance tests that require the vahe to be open.
- b. The purge supply and exhaust isolation valves shall not be opened wider than 33 or 30 degrees, respectively (90 degrees is fully open).
APPLICABillTY: MODES 1, 2, 3, AND 4. ACTION:
- a. With a containment purge supply and/or exhaust isslation valve (s) open for reasons other than given in 3.6.1.7.a above, close the open valve (s) or isolate the penetration (s) within 4 hours, otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With a containment purge supply and/or exhaust isolation valve (s) having a measured leakage rate exceeding the limits of Specification 4.6.1.7.2, restore the inoperable valve (s) to OPERABLE status or isolate the penetrations such that the measured leakage rate does not exceed the limits of Specification 4.6.1.7.2 within 24 hours, otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.6.1.7.1 Each containment purge supply and exhaust isolation valve shall be verified to be sealed closed or open in accordance with Specification 3.6.1.7.a at least once per 31 days. r 4.6.1.7.2 At least once per 6 months, iach containment purge supply and exhaust isolation valve shall be demonstrated OPERABLE by verifying that the measured leakage rate is less than or equal to 0.05 L, when pressurized to Pa' 4.6.1.7.3 At least once per 18 months, the mechanical stop on each containment purge supply and exhaust isolation valve shall be verified to be in place and that the valves will open no more than 33 or 30 degrees, respectiva1y. TURKEY POINI - UNITS 3 & 4 3/4 6-11 AMENDMENT NOS.137 AND 132
CONTAINMENT SYSTEMS 3/4.6.2 DrPRESSURIZATION AND COOLING SYSTEMS CONTAINMENI SPRAY SYSTEM LIMillNG CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and manually transferring suction to the containment sump via the RHR System. APPLICABILITY: MODES 1, 2, 3, and 4. . ACTION:
- a. With one Containme,.t Spray System ineperable restore the inoperable Spray System to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With two Containment Spray Systems inoperable restore at least one Spray System to OPERABLE status within 1 hour or be in at lea- HOT ST ANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore both Spray Systems to OPERABLE status within 72 hours of initial loss or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTOOWN within the following 30 hours.
SURVElLLANCE REQUIREMENTS 4.6.2.1 Each Containment Spray System shall be demonstrated OPERABl.E:
- a. At ieast once per 31 days by verifyi;,9 that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position and that power is available to flow path components that require power for operation;
- b. By verifying that on recirculation flow, each pump develops the '
indicated dif f er ential pressure, when tested pursuant to Specification 4.0.5:
- Containment Spray ? ump 0241.6 psid while aligned in recirculation mode.
1URKEN POINI - UN.'S 3 & 4 3'4 6-12 AMENDMENT N05.137/ND 132
CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMtNTS (Continued)
- c. At least once per 18 months during shutdown by:
- 1) Verifying that each automatic valve in the flow path actuates to its correct position on a containment spray actuation test signal, and i
- 2) Verifying that each spray pump starts automatically on a contair, ment spray actuation test signal. The manual isolation valves in the spray lines at the containment shall be locked closed for the performance of these tests.
- d. At lesst once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructeu.
4 TURAE) POINT - UNITS 3 & 4 3/4 6-13 AMENDMENT N05.137 AND 132
M r. . ?> fi' r , i: , CONTAINMENT' SYSTEMS:- EMERGENCY CONTAINMENT COOLING SYSTEM f. h LIMITING CONDITION FOR OPERATION 3.6.2.2 -Three emergency containment cooling units shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With one of the above required emergency containment cooling units:
inoperable restore the inoperable cooling unit to OPERABLE status , within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- b. With two or more of the above required emergency containment cof.ing '
enits inoperable, restore at least two cooling units to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Restore all of the above required cooling units to OPERABLE status within. H 72 hours of initial loss or be in at least HOT STANDBY within.the next 6 hours'and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS t
- 4. 6. h. 2 Each emergency containment e:ooling unit shall be demonstrated OPERABLE:
- a. At least'once per 31 days by starting each cooler unit from the !
control-room and verifying that each unit motor reaches the nominal g~ operatin'g current for the test conditions and operates for at least ' 15 minutes. -
- b. .At least once per 18 months by: ,
. 1) Verifying thit each unit starts automatically on a safety injection (SI) test signal, and
~
Ej Verifying a cooling water flow rate of greater than or equal to 2000 gpm to each cooler. p ili
. e 3
tX ,
. TURKEY POINT - IJNITS 3 & 4 3/4 6-14 AMENDMENT N05.'137AND 132 i Ib a ,
11L?zi :L 1
~
e p. fCONTAINMENT SYSTEMS 13/4.'6.~3- EMERGENCY CONTAINMENT FILTERING SYSTEM f' .
; LIMITING CONDITION FOR OPERATION
-3.6.3 Three emergency containment filtering units shall be OPERABLE.
APPLICABILITY: ' MODES 1, 2,-3, and 4. L ACTION:
. With one' emergency containment filtering unit inoperable, restore the W erable
- filter to OPERABLE status within 7 days or be in at least HOT STANDP" . .n the next 6' hours and in COLD SHUTOOWN within the following_30 hours. .
4-SURVEILLANCE REQUIREMENTS 4,6.3 Each emergency containment filtering luilt shall be demonstrated OPERABLE: 3
- a. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal "
adsorbers and verifyiag that the system operates for at least 15 minutes; ,
~-
- b. At least once per 18 months or .(1) af ter any structural maintenance ,
on the HEPA filter or charcoal adsorber housings, or (2.) following l operational exposure of filters to effluents from painting, fire, or L chemical release or (3) after every 720 hours of system operation by: 1)~ Performance of a visual inspection'for foreign material and gasket deterioration, an.d. verifying that the filtering unit s.atisfies the in place penetration and bypass leakage testing y acceptance criteria.of greater than or equal to 99% removal of 00P and halogenated hydrocarbons at the . system flow rate of 37,500 cfm 110%;
- 2) Verifying within 31 days after removal, that a labora' tory analy-sis of a representative etebon sample obtained in accordance with applicable portions of Regulatory Position C.6.b of Regula- -
tory Guide 1.52, Revision 2, March 1978, and performed in accordance with ANSI N-510-1975, meets the acceptance criteria e of greater than 99.9% removal of elemental iodine; and that any charcoal failing to meet this criteria be replaced with charcoal that meets or exceeds the critwia of position C.6a of Regulatory i Guide 1.52, Rev. 2; and
- 3) Verifying a system flow rate of 37,500 cfm 110% and a pressure drop across the HEPA and charcoal filters of less than 6 inches water gauge during system operation when tested in accordance with ANSI N510-1975; ,
4 TURKEYPOINI-UNIT 33&4 3/4 6-15 AMENDMENT N05.137AND 132
f"~^,' <, f , it CONTAINMENT SYSTEMS' a p . SURVEILLANCE REQUIREME'NTS (Continued)- s.
- c. After maintenance affecting flow distribution, by performance of a
, 1 visual inspection and an air distribution test at a system flow rate of 37,500'cfm 110%;
- d. At least once.per 18 months by:
m VerifyinC that the system starts on a Safetp Injection-test; ;
- 1) . ' signal aad;
- 2) Verifying that the filter cooling solenoid valves can be opened by operator action and are opened' automatically on a loss of y flow-signal.
- e. After each complete or partial replacement of a HEPA filter bank, by= I performance visual inspection for foreign material and gasket ,
- deteriora' d by verifying that the filtering unit satisfies the 4 in place N .dtion and bypass leakage testing acceptance criteria.
of greatc. ;aan or equal to 99% removal of DOP-test aerosol while-operating.the system at a flow rate of 37,500 cfm 110%; and
- f. After each complete or partial replacement of a charcoal adsorber bank, by performance of a visual inspection for foreign material and gasket.
m deterioration and.by verifying that theEfiltering unit satisfies the-
- in p' lace penetration and bypass leakage testing acceptance criteria of greater than or equal to 99% removal of halogenated hydrocarbon ,
while operating the system at-a flow rate of 37,500 cfm 110%. ] e a 4 4 5
- 4.'
e n 9 1 I i TURKEY POINT - UNITS 3 & 4 3/4 6-16 AMENDMENT NOS.137 AND 132 a;,
. - em .
a.
; s CONTAINMENT-SYSTEMS:
3/4;6.4 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION - 3.6.4- Each containment isolation valve shall be OPERABLE with isolation times less-than or equal to required isolation times. APPLICABILITY:- MODES 1, 2, 3, and 4.
, ACTION:
*With~one'or.more isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is.open and either:
ai Restore-the inoperable valve (s) to OPERABLE status within 4 hours, or
- b. Isolate each af fected penetration within 4 hours by use of at least one deactivated automatic containment isolation valve secured in the isolation position, or
- c. Isolate each af fected penetration within 4 hours by use of at least one closed manual vali.o ar blind flange, or
. ~
d.- Be' in at least; HOT: STANDBY within the next 6 hours and in' COLD SHUTDOWN within the following 30 hours. . SURVEILLANCE REQUIREMENTS
.+
~
4.6.4.1 The-isolation valves shall be demonstrated OPERABLE prior to returning. the valve to service after maintenance, repair or replacement work is performed
'- 'on the valve or its associated actuator, control or power circuit by. performance of a cycling test, and verification of , isolation time.
~
s .
'* CAUTION: The inoperable isolation valve (s) may be part of a system (s).
Isolating the affected penetration (s) may affect the use of the system (s). Consider the technical specification requirements on the affected system (s), y and act accordingly. h TURKEY POINT - UNITS 3 &.4 3/4 6-17' AMEN 0 MENT N05.137 AND 132
w+ -
- 0;
' CONT AINMENT' SYSTEMS -
SURVEILLANCE-REQUIREMENTS (Continued)
' 4.6.~4.2 Each isolation valve shall be demonstrated OPERABLELduring the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:
oc a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" i isolation valve actuates to.its isolation position; i
- b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position; and- ,
.c.~ Verifying that on a Containment Ventilation Isolation test signal, each purge, exhaust and instrument air bleed valve actuates to its .
isolation position, t
. 4.6.4.3 The isolation time of each power-operated or automatic valve shall be
. determined to be within its limit when tested pursuant to Specification 4.0.5.
i'
+ H 1
t 3/4 6-18' ffURKEY POINT - UNITS 3 & 4 AMENDMENT N05.1'37 AND 132 , 4
( ,y
- c. ~i
[; q CONTAINMENT' SYSTEMS , 3/4.6.5 COMBUSTIBLE GAS CONTROL l ,
. HYDROGEN MONITORS-- .
y ., LIMITING CONDITION FOR OPERATION w 3.6.5 Two independent containment hydrogen monitors shall be OPERABLE. APPLICABILITYi MODES 1 and 2. 1 ACTION: s.- With one-hydrogen monitor inoperable, restore the-inoperable monitor to OPERABLE status within'30 days or be in at least HOT STANDBY within the . next 6 hours.. -j i
- b. With both hydrogen monitors inoperable, restore at least one monitor to l OPERACLE status within 72 hours or be in at least HOT STANDBY within the J next 6 hours. -f 1
y SURVEILLANCE REQUIREMENTS j
'i
- (
,,, /
4.6'5.1
. Eac'.h hydrogen monitor shall be demonstrated OPERABLE by the performance I ot a CHANNEL CHECK at least once per 12 hours, an ANALOG CHANNEL OPERATIONAL L TEST at least once per 31 days, and at least once per 92 days on a STAGGERED TEST BASIS by performing a CHANNEL CALIBRATION using sample gas containing: .d
- ~
J j a.. One . volume percent : hydrogen, balance nitrogen, and
-i b.- Four volume percent hydrogen, balance nitrogen, u
-4.6.5.2 The flow path to each hydrogen monitor. shall be demonstrated LPERABLE at least once per 31 days by a system walkdown to verify that each accessible
.{
=_ manual, power operated, or automatic valve is in its correct position and that 'l power.-is available to those components related~to the operability of th- 1 flowpath, f k
i -p i .) a y
/ -
r e .
. TURKEY P0 INT - UNITS 3 &-4 3/4 6-19 AMEN 0 MENT N05.137AND 132
- w N
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+
y-
- CONTAINMENT SYSTEMS '
If . 3/4.6.6- POST ACCIDENT CONTAINMENT VENT SYSTEM ! y LIMITING CONDITION FOR OPERATION 3.6.6 A Post Accident Containment Vent System shall be OPERABLE. APPLICABILITY- MODES 1 and 2. ACTION: With'the Post Accident Containment Vent System inoperable, restore the Post. Accident-Containment Vent System to 0PERABLE status within 7 days'or be in at
' east HOT STANDBY within 6 hours, i
pt:VEILLANCEREQUIREMENTS 4.6.6 Yhe Post Accident Containment Vent System shall be demonstrated OPERABLE:
- a. - At least once:per 31 days by demonstrating system flow path operability via a system walkdown to verify that each accessible manual. valve is in its correct position. .'
i
- b. At least once per 18 months o'r (1) after any structural maintenance ;
of the HEPA filter or-charcoal adsorber housings, or (2) following operational exposure of filters to effluents from painting, fire, or chemical release in any ventilation zone comm'unicating with the system, or (S) af ter 720 hours of system operation or (4)'after replacement of a filter by: F 1) A visual inspection of the system for foreign materials a'nd ,l gasket deterioration and verifying that the filter system satisfies i the penetration and bypass leakage testing acceptance criteria of less than 1% for DQP and halogenated hydrocarbon tests
- l. conducted at a design flow rate of 55 cfm 110%;
- 2) Verifying, within 31 days af ter removal, that a laboratory analysis of a representative carbon sample performed in accordance with ANSI N510-1975, meets the methyl iodide removal criteria of greater than or equal to 90% and that any charcoal failing to meet the criteria be replaced with charcoal'that. meets or.
exceeds the criteria of Position C.6.a of Regulatory Guide 1.52, Revision 2. 4 9 TURKEY POINT - UNITS 3 & 4 3/4 6-20 AMENDMENT NOS.137 AND 132 l 1
7,_, . . _ N? 5
, ft, 4
[i .
-o Q.'CONTAINMENTSYSTEMS-W\ ,
gj . SURVEILLANCE REQUIREMENTS (Continued). c ' 4 ,
,) _h'. 3
, ci <-At least once per 18 months by:
. 1) Verifying that-the pressure drop acrossLthe combined HEPA filter and charcoal ~adsorber is less than 6 inches Water Gauge at a-flow rate of 55 cfm i 10%,.
"i -
-" ' 4
- 2) Visual-inspection of the system and operation of all valves.. ,
t
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e
? TURKEY POINT'- UNITS 3 & 4 3/4 6-21 ' AMENDMENT N05.137 AND 132
., I
nr 4-i t 3, 4.' 7 ' ' PLANT SYSTEMS: 3/4.7.1 TURBINE CYCLE-- SAFETY VALVES c LIMITING CONDITION FOR OPERATION 3.7.1 1 All main steam line Code safety valves associated with each steam; generator'shall be OPERABLE with lift settings as specified in Table 3.7-2.
=d -APPLICABILITY: MODES 1, 2, and 3.
ACTION:
'. lith (3) reactor coolant loops and associated steam generators in-operation and with one.or more main steam line Code saf ty valves inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours, either the in-operable valve is restored to OPERABLE status or the Power. Range Neutron Flux High Trip-Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUT 00WN within the following 30-hours.
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specification 4.0.5. TURKFY POINT - UNITS 3 & 4 3/4 7-1 AMEN 0 MENT N05.137 AND 132
q , lll$; , ' '
..-v u , .
i. i: ^" TABLE 3;7-1 0 ,
. MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE-5AFETY VALVES DURING THREE LOOP OPERATION l 1, -
j L ., - . . h MAXIMUM' NUMBER OF INOPERABLE MAXIMUM ALLOWABLE POWER. RANGE E -SAFETY. VALVES ON ANY NEUTRON FLUX HIGH SETPOINT?
' OPERATING STEAM Gri'ERATOR
' (PERCENT OF RATED THERMAL POWER) ,
h . 1 -82 L E. , 2 54 . e ; ok 3 27 o u TABLE 3.7-2 l l>, STEAM LINE SAFETY VALVES PER LOOP 1: I . ORIFICE SIZE H VALVE NUMBER LIFT SETTING (*1%)* SQUARE INCHES u
; Loop A Loop B Loop C ,
4 1085 psig
- 1. 'RV1400 RV1405 'RV1410 16 ,
- 2. 'RV1401 RV1406 RV1411 1100 psig ,16
'I 1115-psig 16' a
L 3. RV1402 RV1407 RV1412 l o , [, 4. RV1403 RV1408 RV1413- 1130 psig 16 u 2 -l 1
*Theolift setting pressure shall correspond to ambient conditions-of the valve I I
at nominal operating temperature and pressure. l: , 1' , l l
,y p
3/4 7-2
. TURKEY l POINT - UNITS 3 & 4 AMENDMENT N05.137 AND 132 l
e t q, .
; i
. -. 1 PLANT SYSTEMS-
+
AUXILIARY FEE 0 WATER SYSTEM-LIMITING CONDITION FOR OPERATION 3.7.1.2 Two indepandent auxiliary feedwater trains including 3 pumps as specified in Table 3.7-3 and associated flowpaths shall be OPERABLE. ' APPLICABILITY: MODES 1, 2 and 3-g ACTION:
- 1) With one of t'e two required independent auxiliary feedwater trains inoperable, a*'her restore the inoperable train to an OPERABLE status within 72 hours, or place the affected unit (s) in at least HOT STANDBY within the next 6 hours
- and in HOT SHUTDOWN within the ,
following 6 hours. 2)' With both required auxiliary feedwater trains inoperable, within 2- u hours either restore both trains to an OPERAB E status, or ru tore -l one train to ar. OPERABLE status and follow ACTION statement l'above for the other train. If neither train can be restored to an OPERABLE q status within 2 hours, verify the OPERABILITY of both standby. feed- l water pumps and place the affected unit (s) in at least HOT STANDBY within the next 6 hours
- and in HOT SHUTDOWN within the following 6 ]
hours. _0therwise, initiate corrective action to restore et least one l auxiliary feedwater train to an OPERABLE status as sovr au possib1 and follow' ACTION statement 1 above for the other train. l
. ~3) - With a single auxiliary feedwater pump inoperable,-within 4 hours,-
verify OPERABILITY of two independent auxiliary feedwater trains, or
, follow ACTION statements 1 or 2 above as applicable. Upon.verifica-
' tion of the OPERABILITY of two independent auxiliary feedwater trains, restore the~ inoperable auxiliary feedwater pump to an OPERABLE status within-30 days, or place the operating unit (s) in at least H0T SfANDBY within 6 hours
- and in HOT SHUTOOWN Within.the f'ollowing .
6 hours. The provisions of Specification 3.0.4 are not apt.licable during the 30 day period for the -inoperable auxiiiary feedwater pump; SURVEILLANCE-REQUIREMENTS L c 4.7.1.2.1 The required independent auxiliary feedwater trains snsil be , l- + - demonstrated OPERABLE: .
- a. At least once per 31 days on a STAGGERED TEST BASIS by,:
- 1) Verifying by control panel indication and visual opservation of equipment that each steam turbine-driven pump' operates, for 15 minutes or greater and develops a flow of greater than or L *If. this ACTION applies to both units simultaneously, be in at least HOT ,
. . STANDBY within the next 12 hours and in HOT SHUTDOWN within the.following 5 hours. <
TURKEY POINT - UNITS 3 & 4 3/4 7-3 AMENDMENT N05.137AND132
1 % g, r [^t PLANT SYSTEMS w AUXILIARY FEEDWATER SYSTEM
- y. SURVEILLANCE REQUIREMENTS (Continued)
' equal to 373 gpm to the entrance of the.. steam-generators. MThe provisions of Specification ~4.0.4'are not_ applicable for ent'ry into MODES'2 and 3; gg 2)_ verifying by; control panel indication and visual observation of, equipment that the auxiliary feedwater discharge. valves-and the steam supply and turbine pressure valves operate as required to deliverL the required flow during the pump performance test above;
-3) Verifying that each non-automatic valve in the flow path.that
- is not locked, sealed, or otherwise secured iii position is in its correct position; and
- 4) .Vnrifying that power is available to those components which require power for flow path operability.
- b. At least once per 18 months by:
- 1) Verifying that each automatic valve in the flow path' actuates to its correct position upon receipt of.each Auxiliary Feedwater Actuation 1 test signal, and
- 2) Verifying that each auxiliary feedwater pump receives a start o' signal as designed automatically upon receipt of each Auxiliary.
, ,. Feedvater Actuation test signal. . .
4.7.1.2.2 An auxiliary feedwater flow path to each steam generator-shall be demonstrated OPERABLE following each COLD SHUTDOWN-of greater than-30 days prior to entering MODE 1 by verifying normal flow to each steam generator.
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. y-I g' -' , TURKEY POINT - UNITS 3 & 4 3/4 7-4 AMENDMENT N05.137 AND 132 it
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~ TABLE 3.7-3
~ AUXILIARY FEE 0 WATER SYSTEM OPERABILITY e
- UNIT TRAIN STEAM SUPPLY FLOWPATH(3)'. PUMP DISCHARGE WATER FLOWPATH I3)' i
#- 3 11E SG 3C via MOV-3-1405 Aor'C{2) SG 3A via.CV-3-2816-Lor SG 38-via MOV-3-1404(1) ' SG 3B via CV-3-2817 SG 3C via~CV-3-2818 i
2 SG 3A via MOV-3-1403 8 or C(2) SG'3A via CV-3-2831 ' or SG 3B via MOV-3-1404(1) - SG 3B via CV-3-2832' SG 3C via CV-3-2833 f 4 1 SG 4C via MOV-4-1405 SG 4A via CV-4-2816 or SG 4B via MOV-4-1404(1) .A or C(2)
~
SG 40 via CV-4-2817 SG 40 via CV-4-2818
~ 4. 2 SG 4A via MOV 4-1403 3g 4A via'CV-4-2831 or SG 4B via MOV-4-1404(1). B' or SG 4B.via C(2)
CV-4-2832 SG 4C via CV-4-2833 NOTES: (1) Steam admission valves MOV-3-1404 and MOV-4-1404 can be aligned to either l train (but not~both) to restore OPERABILITY in the event MOV-3-1403 or l MOV-3-1405, or MOV-4-1403 or MOV-4-1405 are inoperable. . 4 (2)Duringsingleandtwounitoperation,onepumpshallbeOPERABLkineach.
' train and the third auxiliary feedwater pump shall be OPERABl.E and capable of being powered ~ from, and supplying water-to 'either' train, except .as noted
.in ACTION 3 of Technical Sptcification 3.7.1.2. The = third auxiliary feedwater pump (normally the "C" pump) can be aligned to either train to restore'0PERABILITY in'the event one of the required pumps is inoperable,
, J3)1f any local manual realig*1 ment of valves is required when'operat'ing the auxiiiary feedwater pumps, a dedicated individual, who is in communication with the control room, shall be stationed at the auxiliary f.eedwater pump area. .Upon instructions from the control room, this operator would realign i the valves in the AFW system train to its normal operational alignment.
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TURKEY POINT,- UNITS 3 & 4 3/4 7-5 AMENDMENT N05.137AND132
7,
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t-l PLANT SYSTEMS. t CONDENSATE STORAGE TANK LIMITING' CONDITION FOR'0PERATION p> 3.7.1.3 The condensate storage tanks (CST) system shall be OPERABLE with:- , 9
' Opposite Unit in MODES 4, 5 or 6
~
]
An. indicated water volume of 185,000 gallons .in either'or both condensate-storage tanks. . Opposite Unit in MODES 1, 2 or 3 An indicated wate- *olume of 370,000 gallons.
> APPLICABILITY: MODES ?4 i 3.
ACTION: Opposite Unit in MODES 4, 5 or_6 < With the CST system inoperable, within 4 hours restore the CST s'ystem to OPERABLE status or be in at least HOT STANDBY in the next 6 hours and in H0T SHUTDOWN. - within the following 6 hours. Oppcsite Unit in MODES l', 2 or 3 ,
-1)~ With the CST system inoperable due to containing less than 370,000 gallons, but-greater than or equal to 185,000 gallons within 4 hours restore the inoperable CST sys. tem to OPERABLE status or place one unit in at least- .
HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following -) 16 hours.-
~
, 2) With the CST system inoperable with less than 185,000 gallons within l
'I hour restore the CST system to OPERABLE status or be in at'least HOT
~
. STANDBY within the next 12 hours and in HOT SHUTDOWN within*the following ,
6 hours. This ACTION applies to both units simultaneously. *
+
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r A- '* c i -TURKEY POINT - UNITS 3 & 4 3/4 7-6 AMENDMENT N05.13MND 132 I ^ .
4 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Cor.inued) 4.7.1.3 The condensate storage tank (CST) system shall be demonstrated OPERABLE at least once per 12 hours by verifying the contained water volume is within its limit when the tank is the supply source for the auxiliary feedwater pumps. i a . M
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TURKEY POINT - UN)15 3 & 4 3/4 7-7 AMENDMENT N05.137 AND 132 m-
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- PLANT SYSTEMS' SPECIFIC ACTIVITY' a
LIMITING CONDITION FOR OPERATION _ 3.7.1.41 The specific activity of tht Secondary Coolant System shall be less : than'or equal to 0.10 microcurie / gram DOSE EQUIVALENT-I-131. ' APPLICABILITY: MODES 1, 2', 3, and 4. ACTION: With he specific activity of the Secondary Coolant System greater than 0.10 microcurie / gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 ' hours and in COLD SHUTDOWN within the following 30 hours. i l SURVEILLANCE REQUIREMENTS _ 4.7.1.4 The specific activity of the Secondary Coolant System shall be determined to be within the-limit by performance of the sampling and analysis program of Table 4.7,,1.- 4 - s b O
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. i TURKEY POINT - UNITS 3 & 4 3/4 7-8 AMENDMENT NOS.137 AND 132 K
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{>I;> .>, j L,2 TABLE 4.7-1 ('IJ SECONDARY COOLANT' SYSTEM SPECIFIC ACTIVITY-
, SAMPLE AND ANALYSIS PROGRAM
' TYPE OF MEASUREMENT SAMPLE AND' ANALYSIS AND ANALYSIS FREQUENCY
- 1. Gross _ Radioactivity At least once per 72 hours.
Determination
'2. Isotopic Analysis for DOSE 'a) Once per 31 days,-when-S' EQUlVALENT I-131 Concentration- ever the gross radio- '
activity determination-indicates: concentrations greater than 10% of the-allowable limit foi radiciodines. b) Once per 6 months, when-ever the gross' radio-activity determination indicates concentrations
~
less than or equal-to 10% of the allowable limit
.fdf radioiodines.. .
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=!' sTURKEY POINT - UNITS 3 & 4 3/4 7-9 AMENDMENT N05,.137 ND 132 nr .
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MAIN' STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION
- ; 3.7.1. 5 Each' main steam line isolation valve.(MSIV) shall be'0PERABLE.
APPLICABILITY: . MODES 1, 2, and 3. ACTION: MODE'1:: With one MSIV: inoperable but open, POWER' 0PER^, TION -may continue
~
provided the inoperable valve is restored to OPERABLE status within= 24 hours; otherwise be in HOT STANDBY within the next 6 hours and ~j u in' HOT: SHUTDOWN within the following 6 hours. { MODES 2 and 3: With one MSIV inoperable, subsequent operation in MODE 2 or 3 may proceed: provided the isolation valve is maintained closed. -Otherwise, be in HOT f STANDBY within the next 6 hours and in HOT SHUTDOWH within the following 6' hours, i [
]
a SURVEILLANCE REQUIREMENTS _ -l i
't j
4.7.1.5 Each MSIV.shall be demonstrated OPERABLE by verifying fbil closure ~ l within 5 seconds when tested pursuant'to Specification.4.0.5. The provisions of Specification-4.0.4 are not applicable for entry into MODE.3. l -
]
4 E < TURKEY POINT - UNITS 3 & 4 3/4 7-10 AMENDMENT N05.137 AND132 1 A
t PLANT-SYSTEMS-
=$_TANDBY FEEDWATER SYSTEM s
. LIMITING CONDITION FOR OPERATION' 9 .h i ~3.7.1.6: Two standby feedwater pumps shall be OPERABLE
- and at least 60,000
[ gallons of water (available volume), shall be in the Demineralized Water
' Storage Tank **.
i , APPLICABILITY: MODES 1, 2 and 3 ACTION: la. WIth one standby feedwater. pump inoperable, restore the inoperable pumps to available status within 30 days or submit a SPECIAL REPORT per 3.7.1.6d. b.. With both standby feedwater pumps inoperable:
- 1. WithinL24 hours, notify the NRC and provide cause for inoperability .
and plans to restore pump (s) to OPERABLE status and,
- 2. Submit a SPECIAL REPORT per 3.7.1.6d.
- c. Wii1 less than 60,000 gallons of water in the Demineralized Water Storage Tank restore the available volune to at least 60,000_ gallons within 24 hours or submit a SPECIAL REPORT per 3.7.1.6d.
- d. If a-SPECIAL REPORT is required per the above specificatioi: submit a report describing the cause of the inoperability,' action taken and a
, schedule for restoration within 30 days in accordance with 6.9.2..
SURVEILLANCE REQUIREMENTS 4.7.1.6.-1 The Demineralized Water Storage tank water volume s5allibe deter-mined to be within limits at least once per 24 hours. 4.7.1.6.2 At least monthly ' verify the standby feedwater pumps are OhERABLE by. testing in recirculation on a STAGGERED TEST BASIS. .
, =4.7.'1.6.3 At least once per 18 months, verify,operab'ility of the respective
' standby feedwater pump by powering;from the non-safety grade diesel generators and providing fecdwater to the steam generators, y
*These pumps do not require plant safety related emergency power sources for '
operability and the flowpath is normally isclated.
**The Demineralized Water Storage Tank is non-safety grade.
s ' TURKEY POINT - UNITS 3 & 4 3/4 7-11 AMENDMENT N05,.137 A,ND' 132 _ . n-
i, 2
' PLANT: SYSTEMS.
U 3/4.7.2 COMPONENT COOLING WATER SYSTEM
'I T LIMITING CONDITION L FOR OPERATION ~ -
3.7.2 The. Component Cooling Water System (CCW) 'shall be OPERABLE with:.
- a. :Three CCW pumps, and b.- Two CCW heat exchangers. , 3, e APPLICABILITY: MODES 1, 2, 3, and 4. ,
~
ACTION: ( n, a. 'Withi only_two CCW pumps with independent powe supplies OPERABLE, ~ restore the innperable CCW pump to OPERABLE status within 30 days or + , be in HOT STANDBY within the next 6 hours.and in COLD SHUTDOWN within the following 30 hours. The provisions of Specification 3.0.4 are-not applicable.
. t i b, With only one.CCW pump OPERABLE or sith two CCW pumps OPERABLE but' -i s not.frem independent power supplies, restore two pumps from independent power supplies to OPERABLE status within 72 hours or, b:
in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within. .. , l the-following 30 hours'. ') 1
- c. With less than two CCW heat exchangers OPERABLE, restore two heat L .exchangers to 0PERABLE status within 1 hour or be'in HOT STANDBY -l
-within the next 6 hours and in COLD SHUTDOWN within the following 30 I hours. ,
. . b b
' SURVEILLANCE REQUIREMENTS 1
n . 4.7.2 The Component Cooling Water' System (CCW) shall be demonstrated OPERABLE:
- a. At least:once per 12 hours, by verifying that two heat excharigers and one pump are capable of removing design basi.s heat loads.
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'! - JTURKEY POINT - UNITS 3 & 4 3/4 7-12 AMENDMENT N05.137 AND 132
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. - p-t i j JRVEILLANCEREQUIREMENTS'(Continued)
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. b, At'least"once per 31 days by: -(1) verifying that'each valve-(manual, power-op'erated,- or automatic) servicing safety-related equipment that is not locked,' sealed. or otherwise secured in position is in its. '
correct position,~and (2)' verifying'by,a performance test-the heat-1 exchanger surveillance curves.
~
( p '[ c. At least once 'per 18 months during shut.'dtvn', L by verifying that: Each automatic valve servicing safety-related equipment actuates o i. -1)- to its. correct position on a 51 test signal,.and ~ ~ ,, :t-
~
T* c4 .( 2) Each.; Component Cooling Water System pump. starts automatically 3,1
'on a SI test signal.
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- 3) i Interlocks required for CCW operability are OPERABLE.
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PLANT SYSTEMS 3/4.7.3 INIAKE COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 The intake Cooling Water System (ICW) shall be OPERABLE with:
- a. Three 1CW pump 5, and
- b. Two ICW headers.
APPLICABILITY: MODES 1, 2, 3, and 4. ACTION:
- a. With only two ICW pumps with independent power supplies OPERABLE, restore the inoperable ICW pump to OPERABLE status within 7 days or be in HOT STANDBY within the next 6 hours and in COLD SHU1DOWN within the following 30 hours. The provisions of Specification 3.0.4 are not applicable.
- b. With only one ICW pump OPERABLE or with two ICW pumps OPERABLE but not from independent power supplies, restore two pumps from independent powe supplies to OPERABLE stetus within 72 hours or be in HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
- c. With only one ICW header OPERABLE, restore two headers to OPERABLE status within 72 hours or be in HOT STANDBY within the next 6 beurs and in COLD SHUTOOWN within the following 30 hours.
S1' : LtANCE PEQUIREMENTS _ .-. 4.7.3 The Intake Cooling Water System (ICW) shall be demonstrated OPERABLE:
- a. At least once per 31 day: by verifying that each valve (manual, power-operated, or automatic) servicing safety-related equiprnent that is not locked, sealed, or otherwise secured in position is in its correct position; and
- b. At least once per 18 months during shutdown, by veritying that:
- 1) Each automatic valve servicing safety relat.ed equipmynt actuaten tc it; correct position on a Si test signal, ana
- 2) Each Intake Cooling Weter Systen pJmp Starts automatically on a 51 te;t signal.
- 3) Interlocks required f or system operabili ty at e OPERABt.E.
TURKEY POINT - LNITS 3 & 4 3/4 T It AMENOMENT N05.137 AND 132 [ !.' ll l . ' j f j f. [
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. di 3/417.4 5 UbTIMATEHEATSINK. ,
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, LIMITING CONDITION FOR OPERATION '
'/ ;
h- ,j; . . 3.7.41 The ultimatet heat *. ink shall be OPERABLE with an average supply water temperature to the Intake Cooling Water System less than~ or equal to 100'F. a h ' APPLICABILITY: MODES 1, 2, 3, and 4.' e
- ACT' ION: .,
. rWith the reijuirements ofithe above specification not satisfied, be in at least " -z.- l 2 l '- HOT STANDBY.'within.12-hours and in COLD SHUTDOWN within the foll.owing 30 hours.
This ACTION'shall be applicable ~to' both units. simultaneously. " Q
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SUkVEILLANCEREQUIREMENTS . i
](:.
LD 4.7.4 The ultimate heat sink shall be determi N d OPERABLE at least once per T 24/ hours by.~ verifying-the average supply water temperature" to the Intake i j . Cooling Water System-to be within its limit. l i( . M . ': f,
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$
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/j/ % TEST TARGET (MT-3) l.0 an lua l;jjU1 g
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6" > 4 (#y%,, $ >777 p g - _
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- 4. PLANT SYSTEMS 3/4.7.5 CONTROL ROOM EMERGENCY VENTILATION SYSTEM
. LIMITING CONDITION FOR OPERATION 3.7.5 The Control Room' Emergency Ventilation System shall'be OPERABLE.
APPLICABILITY: All MODES. ACTION: MODES 1, 2, 3 and 4: , With the Control Room Emergency Ventilation System i'noperable, suspend all movement of fuel in the spent fuel pool and restore the inoperable system to OPERABLE status within 84 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. If this ACTION applies to both units simultaneously, be in HOT STANDBY within 12 hours and in COLD SHUTDOWN within the following 30 hours.
-MODES 5 and 6:
With the Contro'. Room Emergency Ventilation System inoperable, suspend all operations involving CORE ALTERATIONS, movement of fuel in the spent fuel pool, or positive reactivity changes. This ACTION shall apply to both units simuitaneously. SURVEILLANCE REQUIREMENTS 4.7.5 The Control Room Emergency Ventilation System shall be demonstrated OPERABLE:
- a. At least once per 12 hcurs by verifying that the control room air temperature is less than or equal to 120 F;
- b. At least once per 31 days by' initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that -
the system operates for at least 15' minutes;
- c. :At least once per 18 months or (1) after 720 hours of system opera-tion, or (2) af ter any structural maintenance on th HEPA filter or e charcoal adsorber housings, or (3) following operat. anal exposure of
\ the filters to effluents from painting, fire, or chemical release in any ventilation zone communicating with the system, or (4) after complete or partial replacement of a filter bank by:
Qi 3/4 7-16 l k TURKEY POINT - UNITS 3 & 4 AMENDMENT N05.137 AND 132 3 w
i b [. g< ,' [. l PLANT SYSTEMS SURVEILLANCE' REQUIREMENTS (Continued) i
- 1) ~ Verifying that the air cleanup system satisfies trie in place.
penetration and bypass leakage testing acceptance criteria of
- greater than or equal to 99% DOP and halogenated hydrocarbon
. removal at a system flow rate of 1000 cfm 110%;
- 2) Verifying, within 31 days af ter removal, that a; laboratory analysis of a representative carbon sample obtained in accordance with-Regulatory Position C.6sb of Regulatory Guide 1.52, Revision 2, i March 1978, and analyzed per ANSI N510-1975, meets the criteria-for methyl iodine removal efficiency of greater than or. equal to.
90% or the charcoal be replaced with charcoal that meets or exceeds the criteria of position C.6 a. of Regulatory Guide 1.52 1 (Revision 2), and: 1
- 3) Verifying by a visual inspection the absence of foreign '
materials and gasket deterioration.
- d. At least once per 12 months by verifying that the pressure drop. j across the combined HEP.6 filters and charcoal adsorber banks is-less than 6 inches Water Gauge while operating the system at a flow rate of 1000 cfm 110%;
e
- e. At least once per 18 months by verifying that on a Containment Phase "A" Isolation test signal the system automatically switches into the- !
recirculation mode of operation. i . I i 1 i i TURKEY POINT - UNITS 3 & 4 3/4 7-17 AMENDMENT N05.137 AND 132
PLANT SYSTEMS 3/A.7.6 SNUBBERS , LIMITING CONDITION FOR OPERATION 3.7.6 All snubbers shall'be OPERABLE. The' only snubbers excluded from the requirements are those installed on nonsafety-related systems and then only-if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system. APPLICABILITY: MODES 1, 2, 3, and 4. MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES. q ACTION: With one or more snubbers inoperable on any system, within 72 hours replace or re-store the inoperable snubber (s) to OPERABLE status and perform an engineering eval- , uation per Specification 4.7.6f. on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system. SURVEILLAN'CE REQUIREMENTS 4.7.6 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program in addition to the require- ' ments of Specification 4.0.5.
, a. Inspection Types As used in this specification, type of snubber shall mean snubbers of.the same design and manufacturer, irrecpective.of capacity.
- b. Visual Inspections Snubbers are categorized as inaccessible or accessible during reactor operation. Each of these groups (inaccessible and accessible) may i be inspected independently according to the schedule.below. The-first inservice visual inspection of each type of snubber shall be performed af ter 4 months but within 10 fronths of conmencing POWER OPERATION and shall include all snubbers. If all snubbers of each type (on any system) are found OPERABLE during the first inservice-visual inspection, the second inservice visual inspection (of that system) shall be performed at the first refueling outage. Otherwise, subsequent visual inspections of a given system shall be performed in 1 accordance with the following schedule:
No. of Inoperable Snubbers of Each Type Subsequent Visual- ; Inspection Period * ** (on any system) per Inspection Period O 18 months i 25% 1 12 months i 25% 2 6 months i 25% 3,4 124 days t 25% 5,6,7 62 days i 25% 8 or more 31 days i 25%
*The inspection interval for each type of snubber (on a given system) shall'not be lengthened more than one step at a time unless a generic problem has been identified and corrected; in that event the inspection interval may be length-ened one step the first time and two steps thereafter if no inoperable snubbers .of that type are found (on that system). **The provisions of Specification 4.0.2 are not applicable.
TURKEY POINT - UNITS 3 & 4 3/4 7-18 AMENDMENT N05.137 AND 132
p PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- c. Visual-Inspection Acceptanc'e Criteria i
Visual inspections shall verify that: (1) there are no visible indi-cations of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are secure, and (3) fasteners for attachment ~of the snubber to the component and to the snubber anchorage are secure. Snubbers which_ appear inoperable as a result of visual s
; inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, provided that: (1) the cause of the rejection is clearly established and remedied for that partic-
, ular snubber and for other snubbers that may be generically susceptible; and (2) the affected snubber is functionally tested in the a5-found 4 condition and determined OPERABLE per Specification 4.7,6e. All )
snubbers connected to an inoperable common hydraulic-fluid reservoir ! I shall be counted as inoperable snubbers. -l l
- d. Functional Tests I i
l For each unit during refueling shutdown, a representative sample of snubbers shall be tested using the following sample plan:
- 1) At least 10% of the total number of safety related snubbers for j the respective unit identified by site records shall be -func- '
tionally tested either in place or in a bench test. For each snubber of a type'that does not'. meet the functional test accep-
- tance criteria of Specification 4.7.6e, an additional 10% of that type of snubber shall be functionally tested until no more ;
failures-are found or until all snubbers of that type have been functionally tested;
- 2) The representative sample selected for functional testing shall .l include the various configurations, operating environments and ;
the range of size and capacity of snubbers. At.least 25% of ! the snubbers in the representative sample shall inclede snubbers ! from- the following categories; , A. Snubbers within 5 feet of heavy equipment (ex. valves, ) pumps, turbines, motors, etc.) B. Snubbers within 10 feet of the discharge from a safety j relief valve.
- 3) Snubbers identified by site records as "Especially Difficult to Remove" or in "High Radiation Zones During Shutdown" shall also -
be included in the representative sample.* ] i
=
- Permanent or other exemptions from functional testing for individual snubbers
- in these categories may be granted by the Commission only if a justifiable basis for exemption is presented and/or snubber life destructive testing was performed to qualify snubber OPERABILITY for all design conditions at either
[ the completion of their fabrication or at a subsequent date. l , TURKEY POINT - UNITS 3 & 4 3/4 7-19 AMENDMENT N05.137 AND 132 h
l P PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Contin'ued)2 In addition to the regular sample, snubbers which failed the previous functional test shall be retested during the next test period. If a spare snubber has been installed in place of a-
.. failed snubber, then both the failed snubber (if it is repaired and installed in another position) and the spare snubber shall' be retested. Test;results of these snubbers may not-be included for the re-sampling. -
- e. Mechanical Snubbers Functional Test Acceptance Criteria >
The snubber functional test shall verify that:
- 1) Activation (restraining action) is achieved with the specified i range of velocity or acceleration in both tension and compression;
- 2) Snubber release rate, where required, is within the specified range in tension and compression, -l 3)' The force required to initiate or_ maintain motion of the snubber is within the specified range in both directions of travel.
- f. Funct-ional Test Failure Analysis An engineering evaluation shall be made of each failure to meet the i' functional test acceptance criteria to determine the cause of the failure. The results of this evaluation shall be used, if applicable, in selecting snubbers to be tested in an effort to_ determine the _j OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.
If any snubber selected for functional testing either fails to activate or fails to move, i.e. , frozen-in place, the cause will be evaluated under the provisions of 10 CFR Part 21. l Should the'results of the evaluation indicate.that the failure was , caused by either manufacturer or design deficiency, further action shall be taken, if needed, based on manufacturer or engineering a recommendations. L For the snubber (s) found inop wab;e, an evaluation shall be performed L on the components to which the inopereble snubbers are attached. The purpose of this evaluation shall be to determine if the components to which the inoperable snubber (s) are attached were adversely affected by the inoperability of the snubber (s) in order to ensure that the component remains capable of meeting the designed service. TURKEY POINT -UNITS 3 & 4 3/4 7-20 AMENDMENT N05.137 AND 132
D
+ e PLANT-SYSTEMS SURVEILLANCE REQUIREMENTS-(Continued)
- g. Snubber Service Life Monitoring Program A record of the service life of each snubber, the date at which the designated service life commences.and the installation and maintenance records on which the designated service life is based I shall be maintained as required by Specification 6.10.3m.
Concurrent with the first inservice visual inspection and during refueling shutdown thereafter, the installation and maintenance records for_each safety related snubber as identified by-site records shall be reviewod to verify that the inc:cated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. If the indicated service life will be
~
). exceeded prior to the next scheduled snubber service life review, the snubber service life shall.be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This re-evaluation, replacement or reconditioning shall tn> indicated in the records.- 3/4 7-21 AMENDMENT N05.137 AND 132 TURKEY POINT - UNITS 3 & 4 m
E #, PLANT SYSTEMS:
< 3/4.7.7- SEALED 500RCE CONTAMINATION LIMITING' CONDITION FOR OPERATION -
3.7.7 Each sealed source containing radioactive material either in excess of
'100 microcuries of beta and/or gamma emitting material or.5 microcuries of alpha.
emitting material.shall be free of greater than or equal to 0.005 microcurie of removable contamination. A_PPLICABILITY: .At all times. ACTION:
- a. With a sealed source having removable contamination in excess of the above limits, immediately withdraw the n ' led source from use and.
either: 1
- 1. Decontaminate and repair the sealed source, or
- 2. Dispose of the sealed source in accordance with Commission- >
Regulations,
~
- b. The provisions of Specification 3.0.3 are not applicable. 2 SURVEILLANCE REQUIREMENTS q
[ 4.7.7.1 Test Requirements - Each sealed source shall be tested for leakage- l and/or contamination by: :
- a. .The licensee, or i
b, Other persons specifically authorized by the Commission or an- ' Agreement State. The test method shall have a detection sensitivity of at least 0.005
' microcurie per test sample. ,
14.7.7.2 Test Frequencies - Each category of taled sources (excluding startup sources and fission detectors previously subjected to core flux) shall ' be tested at the frequency described below. i s
- a. Sources in use - At least once per 6 months for all sealed sources t 1
containing radioactive materials:
- 1) With a half-life greater than 30 days (excluding Hydrogen 3),
and
- 2) In any form other than gas.
TURKEY POINT - UNITS 3 & 4 3/4 7-22 AMENDMENT N05.137 AND132 I
- r. . _ .
4
n - p PLANT SYSTEMS SURVEILLANCE-REQUIREMENTS (Continued) b .- Stored sources not in use - Each sealed source and' fission detector shall be tested prior to.use or transfer to another licensee unless tested within the previous 6 months.- Sealed sources and fission-detectors transferred without-a certificate indicating the last test date shall be tested prior to being placed into use; and
- c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair -l or maintenance to the source. 'l l 4.7.7.3 Reports - A report shall be prepared and submitted to the Commission j on an annual basis.if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removable l contamination.
1 4.7.7.4 A complete inventory of licensed radioactive materials in possession shall be maintained current at all times. !
. ,l L
l I
)
l d i i i k 4 e i TURKEY POINT.- UNITS 3 & 4 3/4 7-23 AMENDMENT N05.137 AND 132 ) l 4 ( 1
b [. v , PLANT SYSTEMS 3/4.7.8 FIRE SUPPRESSION SYSTEMS ,_ FIRE WATER SUPPLY AND 61STRIBUTION SYSTEM E , LIMITING CONDITION FOR OPERATION 3.7.8.1.The fire Water Supply and Distribution System shall be OPERABLE with:
- a. At least two fire suppression pumps, one electric driven and one' diesel driven with their discharge aligned to the fire suppression header,
- b. Two separate water supplies, each with a minimum contained volume of 300,000 gallons, and r
- c. An OPERABLE flow path capable of taking suction from the' Raw Water Tank I and Raw Water Tank 11 and transferring the water through dis-tribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water flow alarm device on each sprinkler or hose standpipe, and the last valve ahead of the deluge valve on each Deluge or Spray System required to be OPERABLE per Specifications 3.7.8.2, 3.7.8.3, and -
3,7,8.4. l APPLICABILITY: At all times. ACTION: 1
- a. With one pump and/or one water supply inoperable, restore the inoper-able equipment to OPERABLE status within 7 days or provide an alter-nate backup pump or supply. The provisions of Specification 3.0.3 are not applicable. This ACTION applies to both units simultaneously,
- b. With the Fire Water Supply and Distribution System otherwise inoper-able, establish a backup fire water capability within 24 hours. This ACTION applies to both units simultaneously.
. TURKEY POINT - UNITS 3 & 4 3/4 7-24 AMENDMENT N05.137 AND 132 i
9 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS
. 4.7.8.1.1 LThe Fire Water Supply and Distribution System shall be demonstrated OPERABLE: y
- c. At least once per 7 days by. verifying the contained water supply. i volume,
- b. At least once per 31 days by starting the electric motor-driven pump and operating it for at least 15 minutes on recirculat. ion flow, 1
- c. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path is in its correct position,
- d. At least once per 12 months by performance of a system flush,
- e. At.least once per,12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel,
- f. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:
'1) Verifying that each automatic valve in the flow path actuates i to its correct position, ,
- 2) -Verifying that the electric-driven pump develops at least 1880
-gpm at a system pressure of 130.2 psig, and the diesel-driven L pump develops at-least 2350 gpm at a system pressure of 130.2 psig, by verifyir.g 3 points on the pump performance curve. l
- 3) Cycling each valve in the flow path that is not tertable during plant operation through at least one complete cycle of full ,'
travel, and
'4) -Verifying that each fire pump starts sequentially to maintain the Fire Water Supply and Distribution System pressure greater than or equal to 125 psig,
- g. At_least'once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association. {
. TURKEY POINT.- UNITS 3 & 4 3/4 7-25 AMENDMENT N05.137 AND 132
L 3, l-PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)- y 4.7.8.1.2 The fire pump' diesel ~ engine shall be demonstrated OPERABLE:
- a. At least once per 31 days by verifying:
- 1) The fuel storage tank contains at least 375 gallons of fuel, and
- 2) The diesel starts from ambient conditions and operates for at
-least 30 minutes on recirculation flow,
- b. At least once per 92 days by verifying that a sample'of' diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D270-1975 is within the acceptable limits specified in Table 1<of ASTM 0975-1977
-when checked for viscosity and water and sediment; and
- c. At least once per 18 months by subjecting the diesel to an.inspec-tion in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.
4.7.8.1.3 The fire pump diesel starting 24-volt battery bank and charger ; shall be demonstrated OPERABLE: j
.a. At least once per.7 days by verifying that:
- 1) The electrolyte level of each battery is above the plates,. and
- 2) The overall battery voltage. is greater than or equal to 24 volts,
- b. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery, and
- c. At least orce per 18 months.by verifying that:
- 1) The batteries, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration, and
- 2) The. battery-to-battery and terminal connections are clean, tight, free of corrosion, and coated with anticorrosion material.
l I TURKEY POINT - UNITS 3 & 4 3/4 7-26 AMENDMENT N05.137 AND132
,;; c .
p PLANT SYSTEMS: SPRAY'AND/0R' SPRINKLER SYSTEMS.- LIMITING CONDITION FOR OPERATION 3.7.8.2' The following Spray and/or Sprinkler Systems shall be OPERABLE:
- a. Fire Zones 47 and 54 - Component Cooling Water Areas
- b. Fire Zones 45 and 55_- Charging Pump Rooms
- c. Fire Zones 79A - North - South Breezeway
- d. Fire Zones 72, 73, 74 and 75 - Emergency Diesel Generator and Day Tank Rooms 1
APPL 1CABILITY: Whenever equipment protected by the Spray / Sprinkler System is required to be OPERABLE. ACTION: i
- a. With one or more of the above required Spray and/or Sprinkler Systems j inoperable within I hour establish a continuous fire watch with i
backup fire suppression equipment. This ACTION applies to both units i simultaneously. {
- b. The provisions of Specification 3.0.3 are not applicable. 1 SURVEILLANCE REQUIREMENTS 1
4.7.8.2 Each of the above required Spray-and/or Sprinkler Systems shall be demonstrated OPERABLE: > a.. At least once per 31 days by verifving that each valve (manual, power- ! operated, or automatic) in the flow path is in its correct position, l
- b. At least once per 12 months by cycling each testable valve in the. W flow path through at least one complete cycle of full travel,
- c. At least once per -18 months:
- 1) By performing a system-functio, al test which includes simulated automatic actuation of the system, and:
a) Verifying that the automatic valves in the flow path actuate to their correct positions on a test signal, and b) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel. TURKEY POINT - UNITS 3 & 4 3/4 7-27 AMENDMENT NOS.137 AND 132
p j, . ,
'i
. ~ i PLANT SYSTEMS r ,
k.
~ SURVEILLANCE REQUIREMENTS (Continued) ,
- 2) By a visual-inspection of the dry pipe spray and. sprinkler-headers to verify their integrity; and
- 3) By a visual inspection of each nozzle's spray area to verify the ~
spray pattern is not obstructed. -
- d. At least once per 3 years by performing an air or water flow test through each open head spray / sprinkler header and verifying each open head spray / sprinkler nozzle is unobstructed.
4 . l c t TURKEY P0 INT - UNITS 3 & 4 3/4 7-28 AMENDMENT N05.137 AND 132
t FIRE HOSE STATIONS t LIMITING CONDITION FOR OPERATION _ ; 3.7.8.3 The fire hose stations given in Table'3.7-4 shall be OPERABLE. APPLICABIllli: Whenever equipment in the areas protected by the fire hose ; stations is required to be OPERABLE. ACTION:
- a. With one or more of the fire hose stations given in Table 3.7-4 ~
inoperable, provide an equivalent capacity fire hose from the nearest equivalent OPERABLE water source. The fire hose shall be of a length of hose cufficient to provide coverage for.the area left unprotected by the inoperable hose station, and shall be stored in a roll at the outlet of the OPERABLE water supply. The above ACTION requirement r shall be accomplished within 1 hour if the inoperable fire hose is the primary means of fire suppression; otherwise route the additional hose within 24 hours. This ACTION applies to both units simultaneously.
- b. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.8.3 Each of the fire hose stations given in Table 3.7-4 shall be demonstrated OPERABLE: .;
- a. At least once per 31 days, by a visual inspection of the fire hose-l stations accessible during plant operations to assure all required equipment is at the station. 4
- b. At least once per 12 months, by: ,
- 1) Visual inspection of the stations not~ accessible during plant operations to assure all required equipment is at the station,
- 2) Removing the hose for inspection and re-racking, and
- 3) Inspecting all garkets and replacing any degraded gaskets in the couplings.
.c. At least once per 3 years, by:
- 1) Partially opening each hose station valve tn verify valve OPERABILITY and no flow blockage, and
- 2) Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above' maximum fire main operating pressure, l
whichever is greater. L l \ L TURKEY i'0 INT - UNITS 3 & 4 3/4 7-29 AMENDMENT N05.137 AND 132
k > y TABLE 3.7-4 FIRE HOSE STATIONS , IDENTIFICATION LOCATION FIRE ZONE HS-0" 1 EL. 18' - East of 4160V SWGR Room on Column 88 HS-0. s2~ EL. 18' - West of 3A Condensate Pump on Pedestal 87 HS-0' -03 EL. 18' - Passageway South of SG Feed Pump Room 83 HS-03-04 EL. 30' - East of 480V Load Center on Column 105 HS-03-05 EL. 30' - South End of Mezzanine Deck 105 HS-03-06 EL. 42' - NW End of Turbine Deck 117 HS-03-07 EL 42' - North of 6A HPFW Heater 117 , HS-03-08 EL. 42' - NW Corner of Entrance to Elevator 79
.[
HS-04-01 EL. 18' - South of 4160V SWGR Room on Column 82 HS-04-02 EL. 18' - Passageway South of SG Feed Pump Room 78 HS-04-03 EL. 30' - East of 480V Load Center at Stairway -105 HS-04-04 EL. 30' - South End of Mezzanine Deck 105 HS-04-05 EL. 42' - West End of Turbine Deck 117 HS-04-06 EL. 42' - East Side of Turbine Deck and North of 117 6A FW Heater HS-04-07 EL 42' - East Side of Turbine Deck and North of 117 4 6B FW Heater HS-04-08 EL. 42' - Southwest Corner of Turbine Deck 117 HS-AB-01 EL. 18' - East-West Passageway at West End 58 HS-AB-02 EL.18' - East-West Passageway at East End 58 HS-AB-03 EL. 18' - North-South Passageway Outside Unit 3 58 , Charging Pump Room HS-AB-04 EL. 50' - Roof of Unit 3 New Fuel Storage Area 118 HS-AB-05' EL. 50' - Roof of Unit 4 New Fuel Storage Area- 118 i TURKEY POINT - UNITS 3 & 4 3/4 7-30 AMENDMENT N05.137 AND132
p PLANT SYSTEMS , FIRE HYDRANTS AND HYDRANT HOSE HOUSES L LIMITING CONDITION'FOR OPERATION ;
- t 3.7.8.4 The fire hydrants and associated hydrant hose houses given in Table 3.7-5 shall be OPERABLE. <
! APPLICABILITY: Whenever equipment in the areas protected by the fire hydrants , is required to be OPERABLE. ACTION:
- a. With one or more of the fire hydrants or associated hydrant hose houses given in-Table 3.7-5 inoperable, within I hour have sufficient additional lengths of 21/2 inch diameter hose located in an Mjacent .
OPERABLE hydrant hose house to provide service to the unpr m cted-area (s) if the inoperable fire hydrant or_ associated hydrant hose house is the primary means of fire suppression; otherwise, provide 4 the additional hose within 24 hours,
- b. The provisions of Specification 3.0.3 and are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.8.4 Each of the fire hydrants and associated hydrant hose houses given in Table 3.7-5 shall be demonstrated OPERABLE:' <
- a. At least once per 31 days, by visual inspection of the hydrant hose house to assure all required equipment is at the hose house,
- b. At least once per 6 months by visually inspecting each fire hydrant and verifying that the hydrant is not damaged, and c, .At least once per 12 months by:
- 1) Conducting a hose hydrostatic' test at a pressure of 150 psig or i at least 50 psig above maximum fire main operating pressure, whichever is greater,
- 2) Inspecting all the gaskets and replacing any degraded gaskets in the couplings, and
- 3) Performing a flow check of each hydrant to verify its OPERABILITY.
1 TURKEY POINT - UNITS 3 & 4 3/4 7-31 AMENDMENT N0Sl37 AND132 l l
{,' l [ t TABLE 3.7 b. r
-FIRE HYDRANTS' i
NUMBER OF-IDENTIFICATION FIRE ZONE LOCATION HYDRANTS [ FH-01 124 NE Corner of Unit 3 near Vehicle Gate 1 into RCA FH-06 NA 'W of Nuclear Maintenance Building 1 FH-07 86 Unit 3 Transformer Area 1 I FH-08 81 Unit 4 Transformer Area 1- ! FH-09 76 Unit 4 Turbine-Generator Area 1 ;, FH-10 77- Unit 4 Condensate Storage Tank Area 2 FH-11 FH-12 NA Unit 4 New Fuel Storage Area 1 FH-13 123 Refueling Water Storage Area. 1 FH-17 NA Nuclear Dry Storage Area 1 FH-16 NA Steam Generator Storage Area J < 4 TOTAL 11 1 b i TURKEY POINT - UNITS 3 & 4 3/4 7-32 AMENDMENT N05.137AND 132
PLANT SYSTEMS r 3/4.7.9 FIRE RATED' ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.7.9 All fire rated assemblies (walls, floor / ceilings, fire barrier penetration seals, and other fire barriers) separating safety-related fire areas or separating. portions of redundant systems important to safe shutdown within a fire area and all sealing devices in' fire rated assembly penetrations (fire doors, fire windows, fire dampers, cable, piping, and ventilation duct .; penetration seals'; Sall be'0PERABLE. ] . APPLICABILITY: At- n times. ACTION:
- a. With one or more of the above required fire rated assemblies and/or sealing devices inoperable, within I hour either establish a continuous fire watch on at least one side of the affected assembly, or verify the OPERABILITY of. fire detectors on at least cne side of the inoperable assembly and establish an hourly fire watch patrol.
- b. The provisions of Specification 3.0.3 are not applicable. l l
SURVEILLANCE REQUIREMENTS 4.7.9.1 At least once per 18 months the above required fire rated assemblies and penetration sealing devices shall be verified OPERABLE by performing a visual inspection of:
- a. The exposed surfaces of.each fire rated assembly,
- b. Each fire window / fire damper and associated hardware, and
- c. At least 10% of each type of sealed penetration. If apparent changes .in appearance or abnormal degradations are found, a visual
. inspection of an additional 10% of each type of sealed penetration shall be made. This inspection process shall continue until a 10%
sample wito no apparent changes in appearance or abnormel degradation is found. Samples shall be selected such that each penetration will be inspected every 15 years. l TURKEY POINT - UNITS 3 & 4 3/4 7-33 AMENDMENT N05.137AND 132
7 (- -- ;
.4 I: 4'
, t PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) k .:
' 4.7'.9.2~ Each of the above required fire doors shall be verified CPERABLE by inspecting the release and closing mechanism and latches at least once per 6 months, and by verifying:
- a. The OPERABILITY of the fire door sup'ervision system for each electrically supervised fire door by performing a TRIP ACTUATING DEVICE OPERATIONAL TEST at least once per 31 days,
- b. That each locked closed fire door is closed at least once per
- 7 days,
- c. That doors with automatic hold-open and release mechanisms are free of obstructions at least once per 24 hours, and a functional ~ test t is performed at least once per 18 months, and
- d. That each unlocked fire door without electrical supervision
.is closed at least once per 24 hours.
i.
- i
( n l TURKEY POINT - UNITS 3 & 4 3/4 7-34 AF 'NT N05.137 AND 132
i
- 3/4l8 ELECTRICAL' POWER SYSTEMS 3/4.8.1 AC SOURCES OPERATING LIMITING CONDITION FOR OPERATION 3.81. 1 As a minimum, the following AC electrical power sources shall be OPERABLE:
- a. TWO 239 KV-4160 volt startup transformers with associated circuits,
- b. TWO diesel generators each with a day and skid-mounted fuel tank containing a minimum volume of 2,000 gallons of. fuel, and
- c. A fuel. storage system containing a minimum volume of-38,000 gallons of fuel and capable of transferring fuel to day tanks via- a fuel transfer pump.
APPLICABILITY: MODES 1,-2, 3, 4
- ACTION:
- a. With either startup transformer inoperable,
- 1) Demonstrate the OPERABILITY of both diesel generators.by performing surveillance requirement 4.8.1.1.2.a.4 separately, for each diesel generator within 24 hours, if the diesel generator has not been suc-cessfully tested within the past 24 hour;. and at least once per 24 '
l hours while the'startup transformer is inoperc51e,
- 2) Notify the NRC within 24-hours of declaring a startup transformer.
inoperable, 4 l 3) Demonstrate the OPERABILITY of at least two cranking diesel generators by performing surveillance requirement 4.8.-l.1.4 within 12 hours. The requirements of specification 3.0.3 do not apply to this ACTION statement,
- 4) Demonstrate the OPERABILITY of the other startup transformer and its . ,
associated circuits by performing surveillance requirement 4.8.1.1.1 within 1 hour and at least once per 24-hours thereafter, and
- 5) a) For the unit with its startup transformer inoperable in MODE 1, restore the inoperable startup transformer to OPERABLE status within 24 hours or reduce THERMAL POWER to < 30% RATED POWER within:
F the next 6 hours. Restore the inoperable startup transformer to OPERABLE status within 30 days or place both units in at ; l i l- least HOT STANDBY within the next 12 hours and in COLD SHUTOOWN within the following 30. hours. l TURKEY POINT - UNITS 3 & 4 3/4 8-1 AMENDMENT N05.137 AND132
(4 1 l ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION ACTION (Continued) b) ' With the unit in MODES 2, 3 or 4,_ restore the inoperable startup transformer to OPERABLE status within 24' hours or place the unit , in at least HOT STANDBY within the next 6 hours and in COLD- ' SHUTDOWN within the following 30 hours.
- b. With either diesel generator inoperable, for reasons other than the performance of surveillance requirement 4.8.1.1.2.c, s
- 1) Demonstrate the OPERABILITY of the remaining diesel generster by performance of surveillance requirement 4.8.1.1.2.a.4 within 24 hours L and once per 24 hours thereafter while the diesel generator is
-inoperable, "
j
- 2) Within 2 hours verify that the engineered safety features that !
depend on the remaining diesel generator are OPERABLE, and verify compliance with specification 3.8.2.1, _
- 3) Demonstrate the OPERABILITY of the startup transformers and their 1 associated circuits by performing surveillance requirement 4.8.1.1.1 l within 1 hour, and at least once per 24 hours thereaf ter, and j g
- 4) Restore the inoperable diesel generator to OPERABLE status within. .
72 hours or be in at least HOT STCDBY within the next 12 hours and in COLD SHUTDOWN within the following 30 hours. This ACTION applies to both units simultaneously, i i c.
~
With'either diesel generator inoperable, for the performance of surveillance requirement 4.8.1.1.2.c, ,
- 1) Demonstrate the OPERABILITY of the remaining diesel generator by l
performance of surveillance requirement 4.8.1.1.2.a.4 within 24 hours E and once per 24 hours thereafter while the diesel generator is inoperable,
- 2) Within 2. hours verify that the engineered safety features that depend on the remaining diesel generator are OPERABLE, and verify l
compliance with specification 3.8.2.1,
'3) Demonstrate the OPERABILITY of the startup transformers and their l-associated circuits by performing surveillance requirement 4.8.1.1.1 within 1 hour, and at least once per 24 hours thereafter, and ,
l 4) Restore the_ inoperable diesel generator to OPERABLE status within J days' or be in at least HOT STANDBY within the next 6 hours and in i l COLD SHUTDOWN within the following 30 hours.
- d. With one startup transformer and one diesel generator inoperable, ,
- 1) Demonstrate the OPERABILITY of the remaining diesel generator by performance of surveillance requirement 4.8.1.1.2.a.4 within 8 hours and once per 24 hours thereafter while the diesel generator is inoperable, TURKEY POINT - UNITS 3 & 4 3/4 8-2 AMENDMENT NOS.137 AND 132 1
t8 t ELECTRICAL-POWER SYSTEMS 1 j
- LIMITING CONDITION FOR OPERATION ACTION (Continued)
- 2) Within 2 hours verify that the engineered safety features that depend on the remaining diesel generator are OPERABLE, and verify
= compliance with specification 3.8.2.1, t
- 3) Demonstrate the OPERABILITY of the remaining startup transformer and its ,
associated circuits by performing surveillance requirement 4.8.1.1.1 within_1 hour, and at least once per 24 hours thereafter, ;
- 4) Demonstrate the~0PERABILITY of at least 2 cranking diesel generators ,
by performing surveillance requirement 4'.8.1.1.4 within 12 hours. The requirements of specification 3,0.3 do not apply to this ACTION. ! statement,
- 5) Comply with the requirements of specification 3.8.1.1 ACTION a.5, ;
and ACTION b.4 or ACTION c.4 whichever is applicable, and ;
- 6) Notify the NRC within 4 hours of declaring both a startup transformer and a diesel generator inoperable.
- e. With two diesel generators inoperable, demonstrate the OPERABILITY of both startup transformers and their associated circuits by perforning
- surveillance requirement 4.8.1.1.1 within 1 hour and, at least once per 24 hours thereafter. Restore at least one of the inoperable diesel generators to OPERABLE status within 2 hours or be in at least HOT STANDBY within the next 12 hours and in COLD SHUTDOWN within the i
following 30' hours. This ACTION applies to both units simultaneously. _, i
- f. With two startup transformers inoperable, >
- 1) Demonstrate the OPERABILITY of both diesel generators by performance I of surveillance requirements 4.8.1.1.2.a.4 within 8 hours and once per 24 holms thereafter while the startup transformer (s) are inoperable, unlets the diesel generators are already operating, l
- 2) Restore one of the inoperable startup transformers to OPERABLE status within 24 hours or place one unit in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. Then place the other unit in at least H0T STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours, and L 3) Notify the NRC within 4 hours of declaring both startup transformers inoperable.
SURVEILLANCE REQUIREMENTS p 4.8.1.1.1 Each required startup transformer and its associated circuits shall be determined OPERABLE at least once per 7 days by verifying correct breaker alignment and indicated power availability. 4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:
- a. In accordance with the frequency specified in Table. 4.8-1, with diesel L
generator surveillances performed nonconcurrently by: TURKEY POINT - UNITS 3 & 4 3/4 8-3 AMENDMENT N05.137 AND 132
4 m ELECTRICAL POWER' SYSTEMS
- SURVEILLANCE REQUIREMENTS (Continued)
- 1) Verifying the-day. and skid-mounted fuel tanks contain a minimum volume of 2,000 gallons of fuel. ,
- 2) . Verifying the minimum fuel volume of 38,000 gallons in the-Diesel Oil Storage Tank. ;
- 3) Verifying that a fuel transfer pump can be-started and transfers fuel from the Diesel Oil-Storage Tank to the Day Tank.
- 4) Verifying.that the diesel generator starts from normal conditions and accelerates to provide 60 + 1.2 Hz frequency and 4160 + 624 volts in < 15. seconds *,
- 5) Verifying that the generator is synchronized, loaded to > 2500 kw within 10 minutes
- and operates for > 60 minutes, and the cooling-system operates within design limits 7
- 6) Verifying that the diesel is aligned to provide standby power to the associated emergency buses,
- b. At least once per 92 days by verifying that a sample of diesel fuel from the Diesel Oil Storage Tank is within acceptable limits when checked for
~iscosity, v water, and. sediment.
l- .
- c. During each Unit 4 refueling outage by:
- 1) Subjecting the diesel to an inspection in conjunction with its manufacturer's recommendations for this class of standby service, i
- d. At least once per 18 months by:
- 1) Verifying the diesel generator's capability to: ,
l a) Reject a load of greater than or equal to 380 kw without l, exceeding-4160 1 624 volts and 60 1 1.2 Hz. L l b) Reject a load of greater than or equal to 2500 kw without tripping. The generator voltage shall return to less than or E equal to 4784 volts within 2 seconds following the load-rejection. , L
*The diesel generator start (15 sec) from normal conditions shall be performed at least once per 184 days-in these surveillance tests. All other engine starts for the purpose of this surveillance testing may be preceded by an engine prelube period and/or other warmup procedures recommended by the manu-facturer so that mechanical stress and wear on the diesel engine is minimized.
TURKEY POINT - UNITS 3 & 4 3/4 8-4 AMENDMENT N05.137 AND 132
l ELECTRICAL POWER' SYSTEMS
^ '
SURVEILLANCE' REQUIREMENTS (Contin v
,- J
- 2) Verifying that diesel generator trips that are made operable during (
l the test mode of diesel operation are inoperable when the diesel is not in the test mode of operation.
- 3) Alternately initiating one of the following two diesel startup tests.- R a). Simulate a safety injection signal, and allow the diesel- ,.
generator to achieve nominal rated voltage and speed. Then simulate a loss of offsite power, and allow the diesel generator ' to load and stabilize. b) . Simulate a loss of offsite power, and allow the diesel generator ; to load and stabilize. Then simulate a safety injection signal, and allow the diesel generator to sequence safety loads and stabilize.
- 4) Monitoring the tests specified in 4.8.1.1.2.d.3 to: i a) Verify proper deenergization and load shedding from the 4160 l volt busses, b) Verify that the diesel generator starts from ambient conditions and accelerates to provide 60 i 1.2 Hz frequency and 4160 1 624 volts in < 15 seconds.
- 5) Verifying that the diesel generator operates for at least 8 hours by
, performing the following tests: 1-a) Load the diesel generator to > 2750 kw during the first 2 hours f of the 8 hour test. During this 2 hour period, increase the load to > 2850' kw until the generator electrical load is stabilized and ! then decrease bar:k to > 2750 kw. b) Load the diesel generator to > 2500 kw during the last 6 hours of the 8 hour test. c) Verify that voltage, frequency, and cooling system functions are within design limits during the 8 hour full-load test.
- 6) Demonstrating the abi.11ty to sequentially:
a) Synchronize the diesel generator with offsite power while the-generator is supplying emergency loads: b) Transfer the emergency load to offsite power; c) Isolate the diesel generator; and d) Return the diesel generator to standby status.
- 7) Verifying the auto-connected loads to each diesel generator do not exceed 2750 kw.
p TURKEY POINT - UNITS 3 & 4 3/4 8-5 AMENDMENT NOS.137 AND 132 1
r L' I
'l l
l 4 ELECTRICAL POWER SYSTEMS
' SURVEILLANCE REQUIREMENTS (Continued)
I
- e. At least once per 10 years or after any modification that could affect diesel generator independence, start both diesel generators simultaneously ,
at a time when both reactors are shutdown and verify that both diesel f generators provide 60 1 1.2 Hz frequency and 41601 624 volts in-less than: or equal to 15 seconds. , 4.8.1.1.3 Reports - All valid diesel generator failures shall be reported to.the Commission in.a Special Report pursuant to Specification 6.9.2 within 30 days. Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the number of failures in the last 100 valid tests is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, L - August 1977. 4.8.1.-l.4 At least two cranking diesel generators shall be demonstrated OPERABLE as required by specification 3.8.1.1 ACTIONS a.3 and d.4 by verifying- I that the cranking diesel generators manually start from normal conditions and accelerate to provide 601 1.2 Hz frequency and 4160t 624 volts and are capable of being aligned to either 4160 volt safety bus. N , 1 t c 1- . l
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7 TURKEY POINT - UNITS 3 & 4 3/4 8-6 AMEN 0 MENT NOS.137 AND ~132
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.b TABLE 4.8-1 p .
-DIESEL GENERATOR TEST SCHEDULE j t y
' J NUMBER OF FAILURES IN' ,
LAST 20 VALID TESTS *- TEST FREQUENCY- ! e. f - <1 Once per 31 days- 3
> Once per 7 days
_2** 1 i s t
- Criteria for detern, ting number of failures and number of valid tests shall
,be in accordance'with Regulatory Position C.2.e of Regulatory Guide 1.108,'
Revision 1, August 1977, but determined on a per diesel ~ generator basis. q I For,the purpose of determining the required valid test frequency, the previous validitest failure count may be reduced to zero if' a complete diesel overhaul to like-new= condition is completed, provided that the overhaul, including appropriate post-maintenance operation and testing, is specifically approved
~by the manufacturer and if acceptable reliability has been demonstrated. The-reliability criterion shall be the successful completion of 14 consecutive >
valid. tests in a single series. Ten of these valid tests shall be in accord-L ance w'ith the' routine Surveillance Requirements 4 8.1.1.2.a.4 and 4.8.1.1.2,a.5; and.four valid tests'in accordance with theE184 day testing requirement.of Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5. If this criterion '
'is not satisfied during the first-series of valid tests, any alternate cri-terion to be used to transvalue the failure count to zero requires prior NRC approval,
**The associated valid test frequency shall be maintained until seven consecu-tive failure free demands have been performed and the number of failures in "
'the last-20 valid demands has been reduced to one.
i t L TURKEY POINT - UNITS 3 & 4 3/4 8-7 AMENDMENT NOS.137 AND 132 i
7
; ELECTRICAL' POWER SYSTEM e
.. AC SOURCES L ll '
p l' SHUTDOWN LIMITING CONDITION FOR OPERATION'
- 3. 8.1. 2 As a minimum, the following AC electricsl power sources shall be !
OPERABLE;
- a. One startup transformer and associated circuits or one offsite circuit .
supplying at least one 4160 vult bus, A or-B, !
- b. One diesel generator with a day and skid-mounted fuel tank containing a minimum volume of 2,000 gallons of fuel, and ]
- c. A fuel storage system containing a minimum volume of 38,000 gallons of q fuel and capable of transferring fuel to day tanks via a fuel transfer q pump.
APPLICABILITY *: MODES 5 and 6 ACTION: ; With less than the above minimum required AC electrical power sources OPERABLE, ! immediately suspend all operations involving CORE ALTERATIONS, positive reac- ! tivity changes, movement of irradiated fuel, or crane operation with loads i over the fuel storage pool, and within 8 hours, depressurize and vent the Reac- , tor Coolant System through a vent greater than or equal to 2.2 square inches, a In addition, when in MODE 5 with the reactor coolant loops not filled, or in : MODE 6 with the water level-less than 23 feet above the reactor vessel flange ! immediately initiate corrective action to restore the required sources to OPER-ABLE status and initiate corrective action to increase RCS inventory as scon l as possible j l SUPVEILLANCE REQUIREMENTS
^
- 4. 8.1. 2 The above required AC electrical power sources shall be demonstrated' OPERABLE by the performance of each of the requirements of Specifications .l 4.8.1.1.2 (except for Specification 4.8.1.1.2.a.5), and 4.8.1.1.3. ]
n
*(Caution - If the opposite unit is in MODES 1, 2, 3 or 4, see Specification 3.8.1.1)
TURKEY POINT - UNITS 3 & 4 3/4 8-8 AMENDMENT N05.137 AND 132 l
c, ,, i/ 1 , p: < o a V :' ' , .; e t 3/4 8.2_ DC SOURCES-3.-
' OPERATING LIMITING CONDITION FOR OPERATION j, 3.8.2.1' As a minimum,. the following DC electrical sources shall be OPERABLE:
- a. 125 volt DC batteries no. 3A,- 3B, 4A and 4B, and
- b. L Battery chargers 3B, 4A and 40 and any two of battery chargers 3A,J4B or.. ,
35. APPLICABILITY: MODES 1, 2, 3, 4 i ' ACTION:
- a. With one of the required batteries inoperable:
1). Within 2. hours verify the OPERABILITY of the opposite train diesel 1 generator or be in at least H0T STANDBY within 12 hours and in~ COLD SHUTDOWN within the following 30 hours. This ACTION applies.to both units simu'ltaneously.
- 2) Restore the inoperable battery to OPERABLE status within 24 hours or be in at least HOT STANDBY within 12 hours and in COLD' SHUTDOWN within the following 30 hours. This ACTION applies-to both units -'
. simultaneously. 1
- b. With one or more of the required battery chargers inoperable, restore one 3 or,more'of the inoperable battery chargers to OPERABLE' status within the
' time limits spe'cified in Table 3.8-1;- otherwise be in'at least HOT STANDBY within the next 12 hours and COLD SHUTDOWN within the following 30 hours. 1 This ACTION applies.to both units simult_neously.
SURVEILLANCE: REQUIREMENTS-i 4.8.2.1 Each 125.vol't battery bank and charger shall be demonstrated OPERABLE:
- a. At least once per 24 hours read and record the pilot cell specific gravity. Tha specific gravity shall be within limits of Table 4.8-2 Category.A.
- b. At least once per 7 days by verifying that: f
- 1) The pilot cell parameters (except specific gravity) in Table 4.8-2 meet the Category A limits, and
- 2) The total battery terminal voltage is greater than or equal to 129 volts on float charge.
TURKEY POINT - UNITS 3 & 4 3/4 8-9 AMENDMENT N05.137 AND 132 ; 1 1 1
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y1
, SURVEILLANCE REQUIREMENTS (Continued)
- c. At least once per 31 days by performing the following:
- 1) Rotate the pilot cell, and
- 2) Check water level and restore as necessary recording amount of water added.
.'{
1 TURKEY POINT - UNITS 3 & 4 3/4 8-10 AMENDMENT N05.137 AND 132 m >.
- 3 L *t f
f TABLE 3.8-1 t l- , BATTERY CHARGER ALLOWABLE OUT-OF-SERVICE TIMES BATTERY CHARGERS (BC) 3B, 4A, and 45 j No-BC's One-BC Two-BC's Three BC's Inoperable Inoperable Inoperable Inoperable i" Battery No-BC's N/A 72 hours 24 hours" I hour Chargers (BC) Inoperable ; 3A 48, and ; 35 One-BC N/A 72 hours
- 1 hour I hour Inoperable .
. Two-BC's 24 hours
- I hour 1 hour I hour Inoperable (
Three-BC's 1 hour 1 hour 1 hour I hour : Inoperable F l
)
i l t
*Within 2 hours verify the OPERABILITY of the two diesel generators, t
TURKEY POINT - UNITS 3 & 4 3/4 8-11 AMENDMENT N05.137 AND 132
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},; ; }
i ; p 3/4 8.2 DC SOURCES- [r SURVEILLANCE REQUIREMENTS (Continued) . 1 [, i
- d. At:1 east once per 92 days and within 7 days after a battery discharge with j j
battery terminal voltage below 105 volts,- or battery overcharge with bat-tery terminal voltage above 143 volts, verify that:- !
- 1) The parameters in Table 4.8-2 meet the Category B limits.'
There is no visible corrosion at either terminals or conne tert or 2) the connection resistance of these items is less inan 150 x 10'g ohm, j and _;
- 3) The average electrolyte temperature of every sixth. cell is above !
60 F. i
- e. At least once per 92 days perform a detailed visual inspection of the battery chargers. j f .' At least onci per 18 months by verifying that: ,
J
- 1) The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,
- 2) The cell-to-cell and terminal connections are clean, tight, and coated with anticorrosion material, n
- 3) The resist 6nce of'each cell-to-cell and terminal connection is less than or equal to 150 x 10 ohm, and
- 4) Each'50 kw battery charger will supply at least 390 1 10 amperes at 125 volts for at least 8 hours and each 37.5 kw battery charger will supply at least 290 1 10 amperes at 125 volts for at least 8 hours.
- g. At least once per 18 months during shutdown by verifying that.the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the '
battery is subjected to a battery service test. L l f
- TURKEY POINT - UNITS 3 & 4 3/4 8-12 AMENDMENT NOS.137 AND 132
. w
r ,, N ( - 9 3 i TABtE 4.8-2 j BATTERY SURVEILLANCE REQUIREMENTS l CATEGORY A (1) CATEGORY B (2) PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE (3) l designated pilot connected cell value for cell connected cell i Electrolyte Greater than Greater than Above top of plates , minimum level minimum level and not overflowing l indication indication mark, and less mark, and less f than k inch than h inch ! above maximum above maximum j level indication level indication mark mark Greater than or Greater than or Greater than or float Voltage equal to equal to equal to 2.1? volts 2.13 volts (6) 2.07 volts j Greater than or Greater than or Not more than 0.020 i equal to equal to below the average of ! Specific 1.200 (5) 1.195 all connected cells . Average of all Average of all j connected cells connected cells , i greater than greater than or 1.205 equal to 1.195(5) 3 I TABLE NOTATIONS
.'(1) for any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours all the Category 0 measurements are taken and found to be within their allowable values, and equalizing charge is started. All Category A and B parameter (s) must ,
berestoredtowithinlimitswithinthenext6 days = . (2) For any Category B parameter (s) outside the limit (s) shown, the battery may be' considered OPERABLE provided that the Category B parameters are within their allowable values, and equalizing charge is started, All - Category B parameter (s) must be restored to within limits within 7 days, (3) Any Category 8 parameter not within its allowable value indicates an inoperable battery. (4) Corrected for electrolyte temperature and level. (5) Or battery charging current is less than 2 amps when on charge. (6)- Corrected for average electrolyte temperature. TURKEY POINT - UNITS 3 & 4 3/4 8-13 AMENDMENT NOS.137 AND132 l
s 4 .f'
. .a ; ,
b . , '4- DC SOURCES [ SHUTDOWN LIMITING CONDITION FOR OPERATION L . 13.8.2.2,'As a minimum, three batteries and associated full-capacity chargers *
.shall'be OPERABLE.
I ' APPLICABILITY **: MODES 5 and 6-ACTION:
-With one or more of.the required 125 volt batteries inoperable and/or associated chargers inoperable, immediately suspend all operations involving CORE ALTERATIONS,
. positive reactivity changes, or movement of irradiated fuel; initiate corrective action to restore the required batteries and associated chargers to OPERABLE status-as soon as possible, and within 8 hours, depressurize and vent the Reactor Coolant System through a vent greater than or equal to 2.2 square inches.
SURVEILLANCE' REQUIREMENTS
'4.8.2.2 The above required 125 volt ba'.tery bank's and associated full capacity chargers shall be demonstrated OPERABLE in accordance with Specification 4.8.2.1.
N
*(defined as a designated charger or a spare charger)
- **(Caution - If the. opposite unit is in MODES 1, 2, 3, or 4, see Specification 3.8.2.1)
TURKEY POINT - UNITS 3 &,4 3/4 8-14 AMENDMENT N05,137 AND 132 e , , , -- - -
y ' l , 3/4 8.3 ONSITE POWER DISTRIBUTION ~ W OPERATING
! LIMITING CONDITION FOR OPERATION 3.8.3.1 The following electrical busses shall be energized in the'specified manner with the 4160 volt and 480 volt Load Center tie breakers open between redundant busses within the unit and between the busses of Units 3 and 4:
- a. One train of AC Busses consisting of:
l 1) 4160 Volt Bus A, t 2) 480 Volt load Center Busses A and C, and
- 3) 480 Volt Motor Control Center Busses A*** and C vital sections, j c b. One train of AC Busses consisting of:
- 1) 4160 Volt Bus B f
,. 2) 480 Volt Load Center Busses B and D, and '
- 3) 480 Volt Motor Control Center Bus B, vital section.
- c. 480 volt Motor Control Center Bus D (vital sectiun)*, ***. :
- d. Opposite unit trains of AC Busses consisting of: ;'
- 1) 4160 Volt Busses, A and B,** and
- 2) Motor Control Centers A, B, and C vital sections l APPLICABILITY: MODES 1, 2, 3, 4 ACTION:
r
- a. With one of the required trains of AC busses not energized, re-energize the train within 8 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. This ACTION does not apply to the Limiting Condition for Operations sections 3.8.3.1.c or 3.8.3.1.d.
- b. With one of the required trains of AC busses of the opposite unit inoperable, ;
for periodic refueling outage maintenance, re-energize the train within 7 - days or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours.
- c. With one of the required trains of AC busses of the opposite unit inoperable, for. reasons other than ACTION b above, re-energize the train within 72 hours or be in at least HOT STANDBY within the next 6 hours and COLD SHUT-i DOWN within the following 30 hours. -
*480 Volt Motor Control Center D is common to Unit 3 and 4.
**0ne.4160 Volt Bus may be de-energized while the 480 Volt Load Centers are crosstied with the associated unit in shutdown modes 5 or 6 and the opposite unit at power upon issuance of an engineering evaluation.
*** Loss of the normal or backup sources of power to Motor Control Centers ,
(MCC) 3A, or D necessitates the application of ACTION f, g, h, or i and does L not imply the inoperability of any single MCC unless both sources of power p are lost concurrently. TURKEY POINT - UNITS 3 &-4 3/4 8-15 AMENDMENT N05.137 AND 132 I < 1
s I l ONSITE POWER DISTRIBUTION
- LIMITING CONDITION FOR OPERATION (Continued) )
I
- d. With.the 480 volt Motor Control Center D vital section not energized, for ,
periodic refueling outage maintenance, verify the OPERABILITY of both die- ! sel generators within I hour and re-energize it within 24 hours or be in at'least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. ;
- e. With the 480 volt Motor Control Center D vital section not energized, for !
reasons other than ACTION ti above, verify the OPERABILITY of both diesel , generators within I hour and re energize it within 8 hours or be in at ! least HOT STANDBY within the next 12 hours and COLD. SHUTDOWN within the following 30 hours. This ACTION applies to both units simultaneously.
- f. With the 480 volt Motor Control Center D vital section normal source of power or backup source of power inoperable, for periodic refueling outage maintenance, verify the OPERABILITY of both diesel generators within 1- ,
-hour and restore the inoperable power supply to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and COLD SHUYOOWN within the following 30 hours,
- g. With the 480 volt Motor Control Center D vital section normal source of power or backup source of power inoperable, for reasons other than ACTION f above, verify the OPERABILITY of both diesel generators within I hour
'and restore the inoperable power supply to OPERABLE status within 72 hours or be in at least HOT STANDBv within the next 12 hours and COLD SHUTDOWN within the following 30 hours. This ACTION applies to both units '
simultaneously,
- h. With the 480 volt Motor Control Center 3A vital section normal source of power inoperable, for periodic refueling outage maintenance, verify the OPERABILITY of both diesel generators within I hour and restore the inoper-able power supply to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and COLD SHUTDOWN within the following 30 hours. ,
i
- i. With the 480 volt Motor Control Center 3A vital section normal source of power inoperable, for reasons other than ACTION h, above, verify the OPERABILITY of both diesel generators within I hour and restore the inoper- q able power supply to OPERABLE status within 72 hours or be in at least HOT 1
STANDBY within the next 12 hours and in COLD SHUTDOWN within the following 30 hours. This ACTION applies to both units simultaneously. SURVEILLANCE REQUIREMENTS I 4.8.3.1' The specific busses shall be determined energized and aligned in the required manner at least once per 7 days by verifying correct breaker alignment . and indicated voltage on the busses. ) TURKEY POINT - UNITS 3 & 4 3/4 8-16 AMENDMENT NOS 137 AND 132
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-ONSITE POWER DISTRIBUTION . " SHUTDOWN i p ,
d LIMITING CONDITION FOR OPERATION l s -i I 3.8.3.2 ;As a minimum, the following electrical busses shall be energized: e,. ,
! One train'of AC emergency busses consisting of one 4160 volt and two
~
, a. :
P.. associated 480 volt AC busse #, .l l ' APPLICABILITY *: MODES 5 and 6 l t;1 " ' g ,
. ACTION:
f With any.of the above required busses not energized, immediately suspend all l operations involving-CORE ALTERATIONS, positive reactivity changes, or movement' !
>- .of-irradiated fuel,. initiate corrective action to energize the required electrical -5 busses as.Soon as possible, and within 8 hours, depressurize and vent the RCS !
through a vent greater.than or equal to'2.2 square inches. ; i 1 4 SURVEILLANCE REQUIREMENTS J 1 4.8.3.2 The specific busses shall be determined energized in the required 1 manner at least once per 7 days by verifying correct breaker alignment and ; indicated voltage on the busses. ; t- 3. l f I P i. , ; l=+ *(Caution - if the opposite unit-is in MODES 1, 2, 3, or 4, see J Specification 3,8.3.1) i y f J-1
. TURKEY POINT - UN115 3 & 4 3/4 8-17 AMENDMENT NOS.137 AND 132 t
_ _ .. -. . . _ , , t
- 3/4.9 REFUEL 1NG OPERATIONS 3/4.9.1 BORON CONCENTRATION
\' LIMITING CONDITION FOR OPERATION 3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions is met; either:
- a. A K,ff of 0.95 or less, or
- b. A boron concentration of greater than or equal to 1950 ppm.
APPLICABILITY: MODE 6.* ACTION: With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 4 gpm of a solution contdning greater than or equal to 20,000 ppm boron or its eativalent until h,f, is reduced to less than or equal to 0.95 or the boron concentration is restored to greater than or equal to 1950 ppm, whichever is tho more restrictive. SURVEILLANCE REQUIREMENTS
- 4. 9.1.1 The more restrictive of the above two reactivity conditions shall be determined prior to:
- a. Removing or unbolting the reactor vessel head, and
- b. Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.
4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling
- canal shall be determined by chemical analysis at least once per 72 hours.
4.9.1.3 Valves isolating unborated water sources ** shall be verified closed and secured in position by mechanical stops or by removal of air or electrical power at least once per 31 days. 4.9.1.4 The spent fuel pit boron concentration shall be determined at least once per 31 days.
*The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the vessel head :losure bolts less than fully tensioned or with the head removed.
**The primary water supply to the boric acid blender may be opened under administrative controls for makeup.
l TURKEY POINT - UNITS 3 & 4 3/4 9-1 AMENDMENT N05.137 AND 132
- _ _ _ _ _ _ _ _ _ _ _ _ _ m
a \? ' !g - REFUELING OPERATIONS l 3/4.9.2 INSTRUMENTATION l o LIMITING CONDITION FOR OPERATION-i F 3.9.2 As a minimum, one primary Source Range Neutron Flux Monitor with con- i l tinuous visual indication in the control room and audible indication in the containment and control room, and one of the remaining three Source Range ; h Neutron Flux Monitors (one primary or one of the two backup monitors) with l continuous visual indication in the control room shall be OPERABLE. ; APPLICABILITY: MODE 6. $ I ACTION:
- a. With one of the above required monitors inoperable or not operating, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes. l Lb. With both of_the above required monitors inoperable or not operating, }
determine the boron concentration of the Reactor Coolant System at ' least once per 12 hours. i SURVEILLANCE REQUIREMENTS , 4.9.2 Each required Source Range Neutron Flux Monitor shall be demonstrateo
- OPERABLE by performance of:
- a. A CHANNEL CHECK at least once per 12 hours, -
- b. An ANALOG CHANNEL OPERATIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and .
- c. An ANALOG CHANNEL OPERATIONAL TEST at least once per 7 days, i
1 TURKEY POINT - UNITS 3 & 4 3/4 9-2 AMENDMENT NOS.137 AND 132
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,' REFUELING CPERATIONS 3/4.9.3 DECAY' TIME e LIMITING CONDITION FOR OPERATION 3.9.3The reactor.Shall be suberitical for at least 100 hours.
APPLICABILITY: During movement of irradiated fuel in thel reactor vessel. ACTION: i With the reactor suberitical for less than 100 hours, suspend all operations involving movement of irradiated fuel in the reactor vessel. t SURVEILLANCE REQUIREMENTS 14.9.' he. reactor'shall be determined to have been suberiticalifor at least-100 t,...rs by verification of the date and time of. suberiticality prior to movement of irradiated fuel in the reactor vessel. 7 i i i
't I
s
-TURKEYLPOINT'- UNITS 3 & 4 - 3/4 9-3 AMENDMENT N05.137 AND 132 T 4
,y 4g kEFUELING OPERATIONS-3/4.9.4- CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION
'3.9.4 The containment building penetrations shall be in the following status:
- a. The equipment door closed and held in place by a minimum of four bolts,
- b. A minimum of one door in each airlock is closed, and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:*
- 1) Closed by an isolation' valve, blind flange, or marual valve, or 2)' Be capable of being closed by an OPERABLE automatic containment ventilation isolation valve.
APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION:. With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building. SURVEILLANCE REQUIREMENTS 4.9.4 Each of the above required cor'tainment building penetrations shall be determined to be either in its closed /is= lated condition or capable of being closed by an OPERABLE automatic containment ventilation isolation valve within !
-100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of irradiated fuel in the containment building by:
- a. Verifying the penetrations are in their closed / isolated condition, or
- b. Testing the containment ventilation isolation valves per the appli-cable portions of Specification 4.6.4.2. j i
- Exception may be taken under Administrative Controls for opening of certain valves and airlocks necessary to perform surveillance or testing requirements.
)
. TURKEY POINT - UNITS 3 & 4 3/4 9-4 AMENDMENT N05.137 AND 132
c, I
'g iL-7 K REFUELING OPERATIONS I 3/4.9.5 COMMUNICATIONS-LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station.
APPLICABILITY:' During CORE ALTERATIONS.- e , i; ACTION: i i: : When direct communications betwt-en the control room and personnel at the
= refueling station cannot be maintained, suspend all CORE ALTERATIONS.-
i
+
SURVEILLANCE REQUIREMENTS
~
4.9.5 Direct communications between the control room and personnel- at th'e refueling station shall be demonstrated within 1 heur prior to the start of and at least once per.12 hours during CORE ALTERATIONS. o . s .
+
a l 7 TURKEY POINT'- UNITS 3 & 4 3/4 9-5 AMEN 0 MENT N05.137 AND 132 ; m
REFUELING OPERATIONS j 1 3/4.9.6 MANIPULATOR CRANE LIMITING CONDITION FOR OPERATION 3.9.6 The manipulator crane and auxiliary hoist shall be used for movement of ; drive' rods or fuel assemblies and shall be OPERABLE with:
- a. The manipulator crane used for movement of fuel assemblies having:
i
- 1) A minimum capacity of 2750 pounds, and-
, 2) An overload cutoff limit less than or equal'to 2700 pounds. i
- b. The auxiliary hoist used for latching and unlatching drive rods having:
1)- A minimum capacity of 610 pounds, and !
- 2) A load indicator which shall be used to prevent lif ting loads _l
? in excess of 600 pounds. APPLICABILITY: During movement of drive rods or fuel assemblies within the reactor vessel.
- ACTION:
With the requirements for crane and/or hoist OPERABILITY not satisfied, suspend use of any inoperable manipulator crane.and/or auxiliary hoist from operations
. involving.the movement of drive rods and fuel assemblies within the reactor vessel.
SURVEILLANCE REQUIREMENTS
, _4.9.6.1 -
At least once each refueling, each manipulator crane used'for movement i of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE-within 100 hours prior to the start of such operations by performing a load test of at least 2750 pounds and demonstrating an automatic load cutoff when the crane load exceeds 2700 pounds. 4.9.6.2 At least once each refueling, each auxiliary hoist and associated load indicator used for movement of drive rods within the reactor vessel shall be demonstrated OPERABLE within 100 hours prior to the start of such operations.by performing a load test of at least 610 pounds.
-TURKEY POINT.- UNITS 3 & 4 3/4 9-6 AMENDMENT N05.137 AND 132 ,
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i: REFUELING OPERATIONS ' 3/4.h.7 CRANE; TRAVEL - SPENT FUEL STORAGE AREAS- , J ; LIMITING CONDITION FOR OPERATION- ! 3.9.7 Loads in excess of 2000 pounds shall be prohibited from travel over i fuel assemblies in the storage pool.* -j APPLICABILITY: With fuel assemblies in.the storage pool. , i i:2 ACTION: i a.- With the requirements of the above specification not' satisfied, place j the crane load in a safe condition. i-p< j p b. 'The provisions:of Specification 3.0.3 are not applicable. ! is L it i
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SURVEILLANCE REQUIREMENTS
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. 1 4.9.7.' Prior to crane operation over fuel assemblies in the spent fuel-storage-pool, verify that each load is 2000 pounds or less.
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- Exception may be taken for the temporary construction crane to be used for "
the re-rack oneration which may be carried over irradiated fuel to facilitate : installation 'of the crane. Lift rigs which meet the design and operational requirements of NUREG-0612 " Control of Heavy Loads at Nuclear Power Plants"
-will"be used while performing this installation.
TURKEY POINT - UNITS'3 & 4 3/4 9-7 AMENDMENT NOS 137AND '32 a . 4
f; l li REFUELING OPERATIONS k 3/4.9.8 ~ RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER QVR LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and
'in operation.*
APPLICABILITY: MODE 6, when the water level above the top of the reactor vessel flange is greater than or equal to 23 feet. ACTION: With no RHR. loop OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours. y: SURVEILLANCE REQUIREMENTS 4.9.8.1.1 At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 3000 gpm at least once per 12' hours. 4.9.8.1.2 The RHR flow indicator shall'be subjected to a CHANNEL CALIBRATION at least once per 18 months.
*The RHR loop may be removed from operation for up to I hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs.
TURKEY POINT - UNITS 3 & 4 3/4 9-8 AMENDMENT N05.137 AND 132
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[' LOW WATER LEVEL , e : L ; LIMITING CONDITION FOR OPERATION l l< l 3.9.8.2' Two. independent residual beat removal (tWR) loops shall' be OPERABLE, i h3 ' and at-least one RHR loop shall be in operation. , n . APPLICABILITY: MODE 6, when the water level above the top of the reactor I L vessel-flange is less than 23 feet. iE i ACTION:
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- a. With less than the required RHR loops OPERABLE, immediately initiate ;
corrective action.to return the required.RHR loops to OPERABLE status, or to establish greater than or equal to 23 feet of water , above the reactor vessel flange, as soon as.possible. !
- b. With no RHR loop in operation, suspend all operations' involving a- :
reduction in boron concentration of the Reactor Coolant System and- 1 immediately initiate corrective action to return the required RHR ; loop to operation. Close all containment penetrations providing a ' direct access.from the containment atmosphere to the outside atmosphere within 4 hours. ! < SURVEILLANCE REQUIREMENTS 4,9.8.2 At.least one RHR loop shall be verified in operation and circulating reactor coolantRat a flow rate of greater than or equal to 3000 gpm at least
.once per 12 hours.
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. TURKEY POINT - UNITS 3 & 4 -3/4 9-9 AMENDMENT N05.137 AND 132
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- REFUELING OPERATIONS
, 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM -
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l l+ : LIMITING CONDITION FOR OPERATION i t 3.9.9 The Containment Ventilation Isolation System shall be OPERABLE. - APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: a.- .With'the Containment Ventilation Isolation System inoperable, close I
- i. each of the containment ventilation penetrations providing direct !
access from the containment atmosphere to the outside atmosphere,
- b. The provisions of Specification 3.0.3 are not applicable. j I
SURVEILLANCE REQUIREMENTS : 4.9.9 The Containment Ventilation Isolation System shall be demonstrated . . OPERABLE within 100 hours prior to the start of and at least once per 7 days-
'during' CORE ALTERATIONS by verifying that Containment Ventilation Isolation i occurs on a High Pidiation test signal from each of the containment radiation monitoring. instrumentation channels.
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-TURKEY POINT - UNITS 3 & 4 3/4 9-10 AMENDMENT NOS.137 AND 132
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< 3/4.9.10 = WATER LEVEL - REACTOR VESSEL
. LIMITING CONDITION FOR OPERATION
+ 3.9.10 - At least 23 feet of water shall be maintained over the top of'the
[ reactor vessel flange. 1,- ' O < APPLICABILITY:' During movement of fuel assemblies or control rods within the containment when either the fuel assemblies being moved or the fuel assemblies a seated within the reactor vessel ~are irradiated while in MODE 6. ACTION: With the requirements-of the above specification not satisfied, suspend all operations involving movement of fuel assemblies or control rods within the l reactor vessel. p , SURVEILLANCE W HPEMENTS- _ 4.9.10 The water level shall be determined to be at least its minimum required
. depth within 2 hours prior to the start of and at least once per 24 hours thereaf ter during movement' of fuel assemblies or control: rods.
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'J TURKEY POINT - UNITS.3 & 4 3/4 9-11 AMENDMENT N05.137 AND 132 i
y I REFUELING OPERATIONS , 3/4.9.11 WATER LEVEL - STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 The water level shall be maintained greater than or equal to elevation 56' - 10"* the spent fuel storage pool.** 3 APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool. ! ACTION:
- a. With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within i its limit within 4 hours, b.- The provisions of Specif.ication 3.0.3 are not applicable, SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.
*During spent fuel rerack operation, the water level may be lowered to a level justified by an engineering safety evaluation. There will be no movement of fuel assemblies with water level lower than 56'- 10" elevation during rerack operation. . **The requirements of this specification may be suspended for more than 4 hours hours ta perform maint w nca provided a safety evaluation is prepared prior to suspensien of the above requirement and all movement of fuel assemblies and '
crane operation with loads in the fuel storage areas are suspended. If the level is nat restored within 7 days, the NRC shall be notified within the next 24 hours. TURKEY POINT - UNITS 3 & 4 3/4 9-12 AMENDMENT N05.137 AND 132
REFUELING OPERATIONS 3/4.9.12 HANDLING OF SPENT FUEL' CASK LIMITING CONDITION FOR OPERATION 3.9.12 The handling of spent fuel cask shall be limited to the following conditions: I
- 1) The spent fuel cask shall not be moved into the spent fuel pit until all the spent ~ fuel in the pit has decayed for a minimum of one thousand five hundred twenty-five (1,525) hours.*
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- 2) Only a single element cask may be moved into the spent fuel pit.
- 3) A fuel assembly shall not be removed from the spent fuel pit in a shipping cask until it has decayed for a minimum of one hundred &
twenty (120) days. - APPLICABILITY: During movement of spent fuel cask in the spent fuel storage area. ACTION: With the requirement of the above specification not satisfied, suspend all movement of the spent fuel cask within the spent fuel storage area. SURVEILLANCE REQUIREMENTS 4.9.12.1 The following required decay times of the spent fuel assemblies shall be determined prior to the movement of a spent fuel cask by verification of date and time the spent fuel assemblies were placed into the spent fuel pit:
- a. 1525 hours of decay of all spent fuel assemblies in the spent fuel pit for movement of a spent fuel cask into the spent fuel pit,
- b. 120 days of decay of the spent fuel assembly in the spent fuel cask prior to removal of the spent fuel cask from the spent fuel pit.
4.9.12.2 Prior to any operations involving spent fuel cask movement into the o spent fuel pit, verify only a single element cask will be moved into the spent , fuel pit. 4.8.12.3 The spent' fuel cask crane interlock shall be demonstrated OPERABLE within 7 days of crane operation and at least once per 7 days (7 days is maximum time between tests; specification 4.0.2 does not apply here) when the crane is , being used to maneuver the spent fuel cask.
- 1he spent fuel cask can be moved into the Unit 4 spent fuel pit af ter a minimum decay of 1000 hours until the new two-region high density spent fuel racks are installed.
TURKEY POINT - UNITS 3 & 4 3/4 9-13 AMENDMENT N05.137 AND 132 ) e
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REFUELING OPERATIONS 3/4.9.13 RADIATION MONITORING LIMITING CONDITION FOR OPERATION 3.9.13 The Containment Radiation monitors which initiate containment and control room ventilation isolation shall be OPERABLE. APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: - a) With one or both radiation monitors inoperable, operation may con-tinue provided the containment ventilation isolation valves are main-tained closed. b) With one or both radiation monitors inoperable, within 1 hour isolate the Control Rcom Emergency Ventilation System and initiate operation of the Control Room Emergency Ventilation System in the recirculation mode. SURVEILLANCE REQUIREMENTS 4.9.13 Each Containment Radiation monitor shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-3. u i TURKEY POINT - UNITS 3 & 4 3/4 9-14 AMENDMENT NOS 137AND132
i REFUELING OPERATIONS 3/4.9.14 SPENT FUEL STORAGE I LIMITING CONDITION FOR OPERATION 3.9.14 The following conditions shall apply to spent fuel storage: ,
- a. Fuel assemblies containing more than 4.1 weight percent of U-235 shall not be placed in the single region spent fuel storage racks.
After installation of the two-region high density spent fuel racks, ' the maximum enrichment loading for the fuel assemblies in the spent fuel racks shall be 4.5 weight percent of U-235. .
- b. The minimum boron concentration in the Spent Fuel Pit shall be ,
1950 ppm. l c.* Storage.in Region II of the Spent Fuel Pit shall be further restricted by burnup and enrichment limits specified in Table 3.9-1. d.* During the re-racking operation only, fuel that does not meet the ' burnup requirement for normal storage in Region II may be stored in Region II in a checkerboard arrangement (i.e., no fuel stored in adjacentspaces). APPLICABILITY: At all times when fuel is stored in the Spent Fuel Pit. ACTION:
- a. With any of conditions a, c or d not satisfied, suspend movement of additional fuel assemblies into the Spent Fuel Pit and restore the spent fuel storage configuration to within the specified conditions.
- b. With boron concentration in the Spent Fuel Pit less than 1950 ppm, suspend movement of spent fuel in the Spent Fuel Pit and initiate action to restore boron concentration to 1950 ppm or greater.
SURVEILLANCE REQUIREMENTS 4.9.14 The boron concentration of the Spent Fuel Pit shall be verified to be 1950 ppm or greater at least once per. month.
- These requirements are applicable only af ter installation of the new two-region high density spent fuel racks.
TURKEY POINT - UNITS 3 & 4 3/4 9-15 AMENDMENT N05.137 AND 132
m i, TABLE 3.9-1 '[r7 SPENT FUEL BURNUP REQUIREMENTS FOR STORAGE IN REGION II 0F THE SPENT FUEL PIT Initial Discharge Burnup w/o M /MT
- 1. 5 0, 1.75 5.0 2.0 9. 0 2.2 12.0 2.4 14.8 2.6 17.6 ;
- 2. 8 20.1 .
l 3.0 22.6 3.2 25.0 3.4 27.4
- 3. 6 29.6 3.8 31.8 4.0 34.0 4.2 36.1 4.5 39.0 Linear interpolation between two consecutive points will yield conservative results. .
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e e TURKEY POINT - UNITS 3 & 4 3/4 9-16 AMENDMENT N05.137 AND 132 4 i $ -. 'h,
3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHU,TDOWN MARGIN LIMITING CONDITION FOR OPERATIDF__ . a 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided , reactivity equivalent to at least the highest estimated control rod worth is J available for trip insertion from OPERABLE control rod (s). APPLICABILITY: MODE 2. ACTION:
- a. With any full-length control rod not fully inserted and with less 1 than the above reactivity equivalent available for trip insertion, 1 immediately initiate and continue boration at greater than or equal to 4 gpm of a solution containing greater than or equal to 20,000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored,
- b. With all full-length control rods fully inserted and the reactor subcriticai by less than the above reactivity equivalent, immedi-ately initiate and continue boration at greater than or equal to 4 gpm of a solution containing greater than or equal to 20,000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full-length control rod either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each full-length control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of-Specification 3.1.1.1. TURKEY POINT - UNITS 3 & 4 3/4 10-1 AMENDMENT NOS.137 AND132
fucil.TF'TCYf;; a LIONS jf4.): ___ '_lEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The group height, insertion, and power distribution limits of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6. 3.2.1, and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided:
- a. The THERMAL POWER is maintained less than or equal to 85% of RATED THERMAL POWER, and
- b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification 4.10.2.2 below.
APPLICABILITY: MODE 1. ACTION: With any of the limits of Specification 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.1, and 3.2.4 are suspended, either:
- a. Reduce THERMAL POWER sufficient to satisfy the ACTION requirements of Specifications 3.2.2 and 3.2.3, or
- b. Be in HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.10.2.1 -The THERMAL POWER shall be detr. ined to be less than or equal to 85% of RATED THERMAL POWER at loast once per hour during PHYSICS TESTS. 4.10.2.2 The requirement; of the below listed specifications shall be performed at least once per 12 hours during PHYSICS TESTS:
- a. Specifications 4.2.2.1 and 4.2.2.5, and
- b. Specification 4.2.3.3.
TURKEY POINT - UNITS 3 & 4 3/4 10-2 AMENDMENT N05.137 AND 132
SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS
, LIMITING CONDITION'FOR OPERATION i 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5,
,and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided: ,
- a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, >
- b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range channels are set at less than or equal to 25% of RATED ;
THERMAL POWER, and
- c. The Reactor Coolant System lowest operating loop temperature (Tavg) is greater than or equal to 531'F.
APPLICABILITY: H0DE 2. , ACTION:
- a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the Reactor trip breakers.
- b. With a Reactor Coolant System operating loop temperature (Tavg) less than 531'F, restore T,yg to within its limit within 15 minutes or be in at least HOT STANDBY within the next 15 minutes.
1 SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5% of RATED THERMAL POWER at least once per hour during PHYSICS TESTS. 4.10.3.2 Each Intermediate and Power Range channel shall be subiected to an
.ANALOGCHANNELOPERATIONALTESTwithin12hourspriortoinitiatlngPHYSICS TESTS.
4.10.3.3 The Reactor Coolant System temperature (T,yg) shall be determined to be greater than or equal to 531'F at least once per 30 minuten during PHYSICS TESTS. l TURKEY POINT - UNIT 5 3 & 4 3/4 10-3 AMENDMENT NOS.137 AND132 l i l =p
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, SPECIAL TEST: EXCEPTIONS ;
i 3/4.10.4 (This specification number is not used) 1 1 o e e i i s 9 TURKEY POINT - UNITS 3}}