Text

Amends 32 & 13 to Licenses NPF-9 & NPF-17,respectively, Revising Tech Specs Re Operating Limits for Transition to Use of Optimized Fuel Assemblies,Boron Injection Sys & Control Rod Insertion Limits
	ML20084C989
Person / Time
Site:	McGuire, Mcguire
Issue date:	04/20/1984
From:	Adensam E; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20084C991	List: ML20084C989; ML20084D007;
References
	TAC-53319, TAC-53320, NUDOCS 8405010113
	Download: ML20084C989 (99)
	v • d • e

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a WASHINGTON, D. C. 20555

\\; s...../ DUKE POWER COMPANY DOCKET NO. 50-369 McGUIRE NUCLEAR STATION, UNIT 1 APENDMENT TO FACILITY OPERATING LICENSE Amendment No. 32 License No. NPF-9 l 1. The Nuclear Regulatory Comission (the Commission) has found that: A. The application for amendment to the McGuire Nuclear Station, Unit 1 (the facility) Facility Operating License No. NPF-9 filed by the Duke Power Company (licensee) dated December 12, 1983, and supplemented February 20, March 9, and March 23, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Comission; 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 conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-9 is hereby amended to read as follows: O l f 8405010113 840420 PDR ADOCK 05000369 l P PDR L i ~ L

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 32, are hereby incorporated into this license. i' l The licensee shall operate the facility in accordance with the Tech-nical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION H5( Elinor G. Adensam, Chief Licensing Branch No. 4 Division of Licensing

Attachment:

Technical Specification Changes Date of Issuance: APR 201!B4

l / p urog% UNITED STATES E } ).e.q '/k NUCLEAR REGULATORY COMMISSION 3 4 ;', p WASHINGTON, D. C. 20555 q') DUKE POWER COMPANY DOCKET NO. 50-370 McGUIRE NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.13 License No. NPF-17 1. The Nuclear Regulatory Comission (the Comission) has found that: A. The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility) Facility Operating License No. NPF-17 filed by the Duke Power Company (licensee) dated December 12, 1983, and supplemented February 20, March 9, and March 23, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the 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 conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D. The issuance of this license amendment will not be inimical to the comon defense and security or to the health and safety of the public; E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied. 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-17 is hereby amended to read as follows: y

O

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.13, are hereby incorporated into this license.

The licensee shall operate the facility in accordance with the Tech-nical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION 64tN, Elinor G. Adensam, Chief Licensing Branch No. 4 Division of Licensing

Attachment:

Technical Specification Changes Date of Issuance: APR 2 01984

ATTACHMENT TO LICENSE AMENDMENT NO.32 FACILITY OPERATING LICENSE NO. NPF-9 DOCKET NO. 50-369 AND TO LICENSE AMENDMENT N0.13' FACILITY OPERATING LICENSE NO. NPF-17 DOCKET NO. 50-370 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain a vertical line indicating the area of change. The corresponding over-leaf pages are also provided to maintain document completeness. Amended Revised Page Page II III IV V VI VIII VII IX X XVI XV XVII XVIII XXII XXI XXIII 1-5 2-2 2-2a 2-5 2-8 2-7 2-9 2-10 2-11 82-1 B2-2 B2-4 3/4 1-1 3/4 1-2 3/4 1-4 3/4 1-3 3/4 1-5 3/4 1-6 3/4 1-Sa 3/4 1-19 3/4 1-20 3/4 1-22 3/4 1-21 3/4 2-1 ~

ATTACHMENT (Con't) 4 .i Amended Revised 4-f Page Page F i-3/4 2-2 4 3/4 2-3 3/4 2-4 1 3/4 2-5 3/4 2-6 3/4 2-7 3/4 2-8 3/4 2-9 1 3/4 2-10 3/4 2-11 3/4 2-12 3/4 2-13 3/4 2-14 2 3/4 2-15 3/4 2-16 3/4 2-17 3/4 2-18 3/4 2-19 3/4 2-20 3/4 2-21 2 3/4 2-22 l 3/4 2-23 3/4 5-1 3/4 5-2 3/4 5-11 3/4 5-12 i B3/4 1-1 B3/4 1-2 B3/4 2-1 B3/4 2-2 B3/4 2-4 B3/4 2-5 I B3/4 2-6 ) B3/4 4-1 B3/4 4-2 83/4 5-2 B3/4 5-1 4 6-9 6-10 6-11 6-12 6-13 6-14 6-16 6-15 6-20 6-19 6-21 6-22 6-23 6-24 I 6-25 6-26 j i i

1 INDEX DEFINITIONS SECTION PAGE 1.23 PURGE - PURGING................................................ 1-5 1.24 QUADRANT POWER TILT RATI0...................................... 1-5 1.25 RATED THERMAL P0WER............................................ 1-5 1.26 REACTOR BUILDING INTEGRITY..................................... 1-5 1.27 REACTOR TRIP SYSTEM RESPONSE TIME.............................. 1-5 1.28 REPORTABLE EVENT.............................................. 1-5 1.29 SHUTDOWN MARGIN................................................ 1-6 1.30 SITE B0VNDARY.................................................. 1-6

1. 31 S LAV E R E LAY T E ST...............................................

1-6 1.32 SOLIDIFICATION................................................. 1-6 1.33 SOURCE CHECK................................................... 1-6 ' i

1. 34 STAGG E R ED T E S T B AS I S...........................................

1-6

1. 3 5 TH E RMA L P0W E R..................................................

1-6 1.36 TRIP ACTUATING DEVICE OPERATIONAL TEST......................... 1-7 1.37 UNIDENTIFIED LEAKAGE........................................... 1-7 1.38 UNRESTRICTED AREA.............................................. 1-7 1.39 VENTILATION EXHAUST TREATMENT SYSTEM........................... 1-7 1.40 VENTING........................................................ 1-7 1.41 WASTE GAS HOLDUP SYSTEM........................................ 1-7 TABLE 1.1, FREQUENCY N0TATION....................................... 1-8 TABLE 1.2, OPERATIONAL M0 DES........................................ 1-9 t I McGUIRE - UNITS 1 and 2 II Amendment No. 32 (Unit 1) j Amendment No.13 (Unit 2) m

4 l INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l SECTION PAGE l], 2.1 SAFETY LIMITS l 2.1.1 REACTOR C0RE................................................. 2-1 ] 2.1.2 REACTOR COO LANT SYSTEM PRESSURE.............................. 2-1 I FIGURE 2.1-la UNIT 1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN f 0PERATION............................................. 2-2 FIGURE 2.1-lb UNIT 2 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION............................................. 2-2a i j FIGURE 2.1-2 (BLANK)............................................... 2-3 I. l~ 2.2 LIMITING SAFETY SYSTEM SETTINGS 1 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS............... 2-4 i l TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS..... 2-5 4 3 1 j BASES j SECTION PAGE 1-i' j 2.1 SAFETY LIMITS 4-[ 2.1.1 REACTOR C0RE................................................. B 2-1 i 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............................. B 2-2 4 l 2.2 LIMITING SAFETY SYSTEM SETTINGS i l 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS................ B 2-3 i i 1 1 1 McGUIRE - UNITS 1 and 2 III Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2) 2 i 1 ) -... -,..~., -.,_,_.. _._..,- -, _ - - -, n.,,, .n ,J,,_,---a -n-.- .,.--.r,,,,,,,n

l l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY................................................ 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL I Shutdown Margin - T,yg > 200*F............................ 3/4 1-1 Shutdown Margin - T,yg 1 200* F........................... 3/4 1-3 Moderator Temperature Coefficient......................... 3/4 1-4 FIGURE 3.1-0 MODERATOR TEMPERATURE COEFFICIENT VS POWER LEVEL UNIT 1................................................ 3/4 1-Sa Minimum Temperature for Criticality....................... 3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Path - Shutdown...................................... 3/4 1-7 Flow Paths - Operating.................................... 3/4 1-8 Charging Pump - Shutdown.................................. 3/4 1-9 Charging Pumps - Operating................................ 3/4 1-10 Borated Water Source - Shutdown........................... 3/4 1-11 Borated Water Sources - Operating......................... 3/4 1-12 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Height.............................................. 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00..................... 3/4 1-16 Position Indication Systems - Operating................... 3/4 1-17 Position Indication System - Shutdown..................... 3/4 1-18 Rod Drop Time (Units 1 and 2)............................. 3/4 1-19 Shutdown Rod Insertion Limit.............................. 3/4 1-20 McQUIRE - UNITS 1 and 2 IV Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

) INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 1 l j SECTION PAGE i Control Rod Insertion Limits.............................. 3/4 1-21 FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER 2 FOUR LOOP 0PERATION................................... 3.4 1-22 f i FIGURE 3.1-2 (BLANK)............................................... 3/4 1-23 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (Unit 1)............................ 3/4 2-1 AXIAL FLUX DIFFERENCE (Unit 2)............................ 3/4 2-2 FIGURE 3.2-la AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER (Unit 1)....................... 3/4 2-4 l [ FIGURE 3.2-lb AXIAL FLUX OIFFERENCE LIMITS AS A FUNCTION OF j RATED THERMAL POWER (Unit 2)....................... 3/4 2-5 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (Z)...................... 3/4 2-6 9 i FIGURE 3.2-2a K(Z) - NORMALIZED F (Z) AS A FUNCTION OF CORE HEIGHT (Unit 1)........ 0.................................. 3/4 f-12 l 1 ] FIGURE 3.2-2b K(Z) - NORMALIZE 0 FQ(Z) AS A FUNCTION OF CORE HEIGHT j (Unit 2)........................................... 3/4 2-13 j 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT 0R.................................................. 3/4 2-14 j FIGURE 3.2-3a RCS TOTAL FLOW RATE VERSUS R (Unit 1)................ 3/4 2-16 i FIGURE 3.2-3b RCS FLOW RATE VERSUS Rg AND R2 - FOUR LOOPS j IN OPERATION (Unit 2).............................. 3/4 2-17 l FIGURE 3.2-4 ROD BOW PENALTY AS A FUNCTION OF BURNUP (Unit 2)..... 3/4 2-18 3/4.2.4 QUADRANT POWER TILT RATI0................................. 3/4 2-19 3/4.2.5 DNB PARAMETERS............................................ 3/4 2-22 ) TABLE 3.2-1 DNB PARAMETERS....................................... 3/4 2-23 1 i 3/4.3 INSTRUMENTATION i ) 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION....................... 3/4 3-1 i 1 McGUIRE - UNITS 1 and 2 V Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2) 4 i ~

t { INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS t SECTION PAGE TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION.................. 3/4 3-2 ) TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES... 3/4 3-9 TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS....................................... 3/4 3-11 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION......................................... 3/4 3-15 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.................................... 3/4 3-16 TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYS-3 i RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS............... 3/4 3-43 Movable Incore Detectors.................................. 3/4 3-45 Seismic Instrumentation................................... 3/4 3-46 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION................... 3/4 3-47 i TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE 4 REQUIREMENTS....................................... 3/4 3-48 Meteorological Instrumentation............................ 3/4 3-49 1 1 TACLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION............ 3/4 3-50 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION { SURVEILLANCE REQUIREMENTS.......................... 3/4 3-51 i ] Remote Shutdown Instrumentation........................... 3/4 3-52 TABLE 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATION........... 3/4 3-53 I r i McGUIRE - UNITS 1 and 2 VI Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

INDEX (' LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 4.3-6 REMOTE SHUT 00WN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................... 3/4 3-54 Accident Monitoring Instrumentation....................... 3/4 3-55 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.................. 3/4 3-56 j TABLE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................... 3/4 3-57 Fire Detection Instrumentation............................ 3/4 3-58 TABLE 3.3-11 FIRE DETECTION INSTRUMENTATION....................... 3/4 3-60 Radioactive Liquid Effluent Monitoring Instrumentation.... 3/4 3-66 TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION.................................... 3/4 3-67 TABLE 4.3-8 RADIOACTIVE LIQUID EFFLUENT MONITORING 1 INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......... 3/4 3-69 i Radioactive Gaseous Effluent M9nitoring Instrumentation... 3/4 3-71 TABLE 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.................................... 3/4 3-72 TABLE 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......... 3/4 3-75 1 Loose-Part Detection System.............................. 3/4 3-78 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. 3/4 3-79 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Powe r Operation.............................. 3/4 4-1 Hot Standby.............................................. 3/4 4-2 Hot Shutdown............................................. 3/4 4-3 Cold Shutdown - Loops Ff11ed............................. 3/4 4-5 l t VII McCUIRE - UNITS 1 and 2

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE Cold Shutdown - Loops Not Fi11ed......................... 3/4 4-6 3/4.4.2 SAFETY VALVES Shutdown................................................. 3/4 4-7 0perating................................................ 3/4 4-8 3/4.4.3 PRESSURIZER.............................................. 3/4 4-9 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 NLABER OF STEAM GENERATORS TO BE INSPECTED OURING 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 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.... 3/4 4-21 3/4.4.7 CHEMISTRY................................................ 3/4 4-22 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............. 3/4 4-23 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS......................... 3/4 4-24 3/4.4.8 SPECIFIC ACTIVITY.......................................... 3/4 4-25 FIGURE 3.4-1 OOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1 pCi/ gram 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 McGUIRE - UNITS 1 and 2 VIII Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

INDEX I LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS ) SECTION PAGE i 3/4.4.9 PRESSURE / TEMPERATURE LIMITS 1 Reactor Coolant System.................................... 3/4 4-30 FIGURE 3.4-2a UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE l UP TO 10 EFPY......................... 3/4 4-31 FIGURE 3.4-2b UNIT 2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS-APPLICABLE 4 i UP TO 10 EFPY......................... 3/4 4-32 l FIGURE 3.4-3a UNIT 1 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP l TO 10 E9FY............................ 3/4 4-33 FIGURE 3.4-3b UNIT 2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS-APPLICABLE UP TO 10 EPFY............................ 3/4 4-34 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITH0RAWAL SCHE 0VLE.................................. 3/4 4-35 Pressurizer............................................... 3/4 4-36 l Overpressure Protection Systems........................... 3/4 4-37 3/4.4.10 STRUCTURAL INTEGRITY...................................... 3/4 4-39 3/4.5 EMERGEhCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS 1 Cold Leg Injection........................................ 3/4 5-1 1 Uppe r Head Inj ecti on...................................... 3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350'F............................. 3/4 5-5 1 3/4.5.3 ECCS SUBSYSTEMS - T 1 350*F............................. 3/4 5-9 i avg 3/4.5.4 (Deleted].................................................. 3/4.5-11 j 3/4.5.5 REFUELING WATER STORAGE TANK............................... 3/4 5-12 ) i I-McGUIRE - UNITS 1 and 2 IX Amendment No.3g (Unit 1) i Amendment No. IJ (Unit 2)

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS [ SECTION PAGE I 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity..................................... 3/4 6-1 Containment Leakage....................................... 3/4 6-2 TABLE 3.6-1 SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS............. 3/4 6-5 i Containment Air Locks..................................... 3/4 6-10 Internal Pressure......................................... 3/4 6-12 Air Temperature........................................... 3/4 6-13 1 Containment Vessel Structural Integrity................... 3/4 6-14 4 l Reactor Building Structural Integrity..................... 3/4 6-15 Annulus Ventilation System................................ 3/4 6-16 Containment Ventilation Syste'm........................... 3/4 6-18 l 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS f f Containment Spray System.................................. 3/4 6-20 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. 3/4 6-22 TABLE 3.6-2 CONTAINMENT ISOLATION VALVES......................... 3/4 6-24 1 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors......................................... 3/4 6-31 Electric Hydrogen Recombiners............................. 3/4 6-32 Hydrogen Control Distributed Ignition System.............. 3/4 6-33 l 3/4.6.5 ICE CONDENSER i Ice Bed................................................... 3/4 6-34 ( X McGUIRE - UNITS 1 and 2

INDEX (' LIMITING CON 0!TIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE TABLE 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND l ANALYSIS PR0 GRAM................................... 3/4 11-10 Dose - Noble Gases........................................ 3/4 11-13 Dose - Iodine-131 and 133, Tritium, and Radioactive Materials in Particulate Form............................. 3/4 11-14 Gaseous Radwaste Treatment System......................... 3/4 11-15 i I Explosive Gas Mixture..................................... 3/4 11-16 i j Gas Storage Tanks......................................... 3/4 11-17 ) l 3/4.11.3 SOLIO RA010 ACTIVE WASTE................................... 3/4 11-18 { 3/4.11.4 TOTAL 00SE................................................ 3/4 11-20 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING I 3/4.12.1 MONITORING PR0 GRAM....................................... 3/4 12-1 i TABLE 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM........ 3/4 12-3 TABLE 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES........................... 3/4 12-9 i TABLE 4.12-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF ) DETECTION (LL0).................................... 3/4 12-10 i 3/4.12.2 LANO USE CENSU5.......................................... 3/4 12-13 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM....................... 3/4 12-15 \\ i j XV McGUIRE - UNITS 1 and 2 ) l

_ - = 1 i i l l INDEX i l-BASES l SECTION PAGE 3/4.0 APPLICA8ILITY................................................ 8 3/4 0-1 i 3/4.1 REACTIVITY CONTROL SYSTEMS l 3/4.1.1 BORATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2 80 RATION SYSTEMS.......................................... 8 3/4 1-2 l 3/4.1.3 MOVABLE CONTROL ASSEM8 LIES................................ 8 3/4 1-3 ) 3/4.2 POWER DISTRIBUTION LIMITS l 3/4.2.1 AXIAL FLUX 0!FFERENCE..................................... B 3/4 2-1 l 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW i RATE ANO NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR....... B 3/4 2-2 1 i FIGURE 8 3/4.2-1 TYPICAL IN0!CATED AXIAL FLUX OIFFERENCE l VERSUS THERMAL P0WER............................ 8 3/4 2-3 i j 3/4.2.4 QUADRANT POWER TILT RATI0................................. 8 3/4 2-6 l l i j 3/4.2.5 DN8 PARAMETERS............................................ 8 3/4 2-6 l 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP and ENGINEERED SAFETY FEATURES r i ACTUATION SYSTEM INSTRUMENTATION........................ B 3/4 3-1 l 3/4.3.3 MONITORING INSTRUMENTATION................................ B 3/4 3-2 j 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. 8 3/4 3-5 I j l 1 3/4.4 REACTOR COOLANT SYSTEM i 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION............. B 3/4 4-1 l l 3/4.4.2 SAFETY VALVES............................................. B 3/4 4-2 i 3/4.4.3 PRESSURIZER............................................... 8 3/4 4-2 l l 3/4.4.4 RELIEF VALVES............................................. 8 3/4 4 3 j 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-3 I 1 1 McGUIRE - UNITS 1 and 2 XVI Amendment No.32 (Unit 1) l Amendment No.13 (Unit 2)

o \\ i INDEX BASES SECTION PACE 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............................... B 3/4 4-7 TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS (UNIT 1).................. B 3/4 4-9 REACTOR VESSEL TOUGHNESS (UNIT 2).................. B 3/4 4-11 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E > 1 MeV) AS A FUNCTION OF EFFECTIVE FULL POWER YEARS................... B 3/4 4-12 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RT FOR REACTOR VESSELS EXPOSED TO 550*F TEMPENkIURE..................................... B 3/4 4 13 ) 3/4.4.10 STRUCTURAL INTEGRITY...................................... 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 (0aleted]................................................. B 3/4 5-2 3/4.5.5 REFUELING WATER STORAGE TANK.............................. B 3/4 5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT....................................... B 3/4 6 1 3/4.6.2 DEP1ESSURIZATION AND COOLING SYSTEMS...................... B 3/4 6 4 3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L................................... B 3/4 6 4 3/4.6.5 ICE CON 0ENSER............................................. B 3/4 6 5 3/4,7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................. B 3/4 7 1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7 3 McGUIRE - UNITS 1 and 2 XVII Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

,' t ll i ( i l, INDEX { r ( l l BASES l I 1 PAGE l SECTION 3/4.7.3 COMPONENT COOLING WATER SYSTEM............................ 8 3/4 7-3 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM.............................. 8 3/4 7-3 t 3/4.7.5 STANOBY NUCLEAR SERVICE WATER P0NO........................ 8 3/4 7-3 3/4.7.6 CONTROL AREA VENT!LATION SYSTEM........................... 8 3/4 7-4 3/4.7.7 AUXILIARY BUILDING FILTERED VENTILATION EXHAUST SYSTEM.... 8 3/4 7-4 7 3/4.7.8 $NU8BERS.................................................. 8 3/4 7-5 3/4.7.9 SEALE0 SOURCE CONTAMINATION............................... 8 3/4 7-6 3/4.7.10 FIRE SUPPRESSION SYSTEMS.................................. 8 3/4 7 6 3/4.7.11 FIRE BARRIER PENETRATIONS................................. B 3/4 7-7 1 3/4.7.12 AREA TEMPERATURE MONITORING............................... 8 3/4 7-7 3/4.7.13 GROUN0 WATER LEVEL......................................... 8 3/4 7 8 ( l 3/4.8 ELECTR! CAL POWER SYSTEMS 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES. 0.C. SOURCES AND ONSITE POWER 0!STRIBUTION SYSTEMS........................ 8 3/4 6 1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES................... 8 3/4 6 3 3/4.9 REFUEL!NG OPERATIONS 3/4.9.1 80RON CONCENTRATION...................................... 8 3/4 9-1 7 3/4.9.2 INSTRUMENTATION........................................... 8 3/4 9 1 l 3/4.9.3 DECAY TIME................................................ 8 3/4 9 1 i 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS......................... 8 3/4 9 1 3/4.9.5 COMMUNICATIONS............................................ 8 3/4 9 2 3/4.9.6 MAN!PULATOR CRANE......................................... 8 3/4 9 2 l 3/4.9.7 CRANE TRAVEL SPENT FUEL STORAGE P00L SU!L0!NG........... 8 3/4 9 2 3/4.9.8 RESIDUAL NEAT REMOVAL AN0 COOLANT CIRCULATION............. 8 3/4 9 2 l 3/4.9.9 and 3/4.9.10 WATER LEVEL REACTOR VESSEL and l STORAGE P00L............................................ 8 3/4 9 3 XVI!! I

} } ), E ADMIN!$7RATIVE CONTROLS ) (' l' SECTION PAGE i i t i 1 6.1 RESPONS!8!LITY................................................. 6-1 l 6.2 ORGAN!ZAT!0N i 6.2.1 0FFSITE...................................................... 61 i j 6.2.2 UNIT 5TAFF................................................... 6-1 i FIGURE 6.2-1 0FFSITE ORGANIZATION................................. 63 i FIGURE 6.2-2 STATION ORGAN!ZAT!0N................................. 6-4 TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION....................... 6-5 6.2.3 STATION SAFETY REVIEW GROUP (SSRG) l j Function............................................ 6-7 l< Composition............................................... 6-7 { Responsibilities.......................................... 6-7 Authority................................................. 67 I Records................................................... 6-7 6.2.4 SHIFT TECHNICAL ADV!50R...................................... 67 l t i j 6.3 UNIT STAFF QUALIFICATIONS...................................... 67 4 l 6.4 TRAINING....................................................... 67 [ 2 ] l 6,5 REVIEW AND AUDIT i i ~ 6.5.1 TECHNICAL REVIEW AND CONTROL i I j Activities................................................ 68 + J ,l l 4 i XXI McGU!RE - UNITS 1 and 2 l

t i i I 4 1j h INDEX 1 j ADMINISTRATIVE CONTROLS r l \\ SECTION PAGE r l j 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSR8) J j Function.................................................. 6-9 1 Organization.............................................. 6-10 1 i Review.................................................... 6-11 Audits.................................................... 6-11 1 Authority................................................. 6-12 I j Records................................................... 6-13 1 i 1 j 6.6 REPORTA8tE EVENT ACT!0N........................................ 6-13 i l i 6.7 S A F E TY L I M I T V I O LAT I ON......................................... 6-13. r l J j 6.8 PROCEDURES AND PR0GRANS........................................ 6-14 i 6.9 REPORTING REQUIREMENTS ] 6.9.1 ROUTINE REPORT5.............................................. 6-16 j Startup Report............................................ 6-16 ) Annual Reports............................................ 6-17 i j. Annual Radiological Environmental Operating Report........ 6-18 a l Semlannual Radioactive Effluent Release Report............ 6-18 i i i Monthly Operating Reports................................. 6-20 Radial Peaking Factor Limit Report........................ 6-21 b. jr I i 4 4 ) i, NcGUIRE - UNITS 1 and 2 XXII Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

7._ l I i. I i INDEX i 1 ADMINISTRATIVE CONTROLS 1 i SECTION PAGE I. I REPORTING REQUIREMENTS (Continued) i 6.9.2 SPECIAL REP 0RTS.............................................. 6-21 6.10 RECORD RETENTION.............................................. 6-22 i 1 1 6.11 RA0!ATION PROTECTION PR0 GRAM.................................. 6-23 i 4 6.12 HIGH RADIATION AREA........................................... 6-23 ) 6.13 PROCESS CONTROL PROGRAM (PCP)................................. 6-24 4 i 6.14 0FFSITE OOSE CALCULATION MANUAL (00CM)........................ 6-25 1 4 j 6.15 MAJOR CHANGES TO RA010 ACTIVE LIQU!O. GASEQUS. AND SOLIO i WASTE TREATMENT............................................. 6-26 1 1 i I I i i i l t } l- } I I 1 L i i i i s i i i .I i t i [l [ McGUIRE - UNITS 1 and 2 XXI!! Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

i s DEFINITIONS PURCE. rtRGING 1.21 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentra-tion or other cperating condition, in such a manner that replacement air or gas is required to purify the confinement. QUADRANTP0yERTILTRAT!ff 1.24 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 outnut 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 for computing the average. RATED THF9 MAL POWER 1 25 RATED THERMAL POWER shall be a total core heat transfer rate to the reactor coolant cf 3411 MWt. WEACTOR BUILDING INTEGRITY 1.26 REACTOR BUILDING INTEGRITY rhall exist when: a. Eacn door in each access optning is closed except when the access opening is being used for normal transit entry and exit, then at least one door shall be closed, b. The Annulus Ventilation System is in compliance with the requirements of Specification 3.6.1.6, and c. The sealing mechanism associatud with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE. REACTOR TRIP SYSTEM RESPONSE TIH6 1.27 The REACTOR TRIP SYSliH RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage. REPORTABLE EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. t McGUIRE - UNITS 1 and 2 1-5 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

.) l l GEO-pie,p.,too,. 9g,4co gp. . 655 650 l 645 Unacceotable Onention 6d0 ' % o,f 655

4 htt, 625- -

629 ~ I o % *'d ~ e l 615 g 1 ), 9 rr, r 619 I 605 Acceptable 600-08'"" 595 - I j 580 - 585 8. .I .2 .5-4 .5 .6 .7 .8 9 t. 1.1 1.2 POVER freacticn or nominell I FIGURE 2.1-la UNIT 1 a REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION l I McGUIRE - UNITS 1 and 2 2-2 Amendment No. 32 (Unit 1) i Amendment No.13 (Unit 2)

) 665, { Flow Per Loop = 95,500 gpm I 655 24cn 658 i 645 22S " ' '*?* O Operation os f, 640< 655< 658' 2000 Dste 625 1900 w

622<

f fa

,0 615 - 612 s 625 672 -g \\ I~I' Acceptacle Operation 5 g, 555< 550< 575< '738. .1 .2 .5 4 .5 .6 .7 .S . ai 1. 1.1 1.2 ~ POVCR ITsac*.icn of rominei1 FIGURE 2.1-lb UNIT 2 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION McGUIRE - UNITS 1 and 2 2-2a Amendment No.32 (Unit 1) Amendment No.13 (Unit 2) i \\

-g m ~ =-s 2 TABLE 2.2-1 W E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E L Manual Reactor Trip N.A. N.A. w"

2. Power Range, Neutron Flux Low Setpoint - < 25% of RATED Low Setpoint - 1 26% of RATED THERMAL POWER -

THERMAL POWER High Setpoint - < 109% of RATED High Setpoint - 5110% of RATED THERMAL POWER THERMAL POWER

3. Power Range, Neutron Flux,

< 5% of RATED THERMAL POWER with < 5.5% of RATED THERMAL POWER High Positive Rate a time constant 1 2 seconds with a time constant 1 2 seconds

4. Power Range, Neutron Flux, 5 5% of RATED THERMAL POWER with 5 5.5% of RATED THERMAL POWER High Negative Rate a time constant 1 2 seconds with a time constant 1 2 seconds m

u.

5. Intermediate Range, Neutron 5 25% of RATED THERMAL POWER 5 30% of RATED THERMAL POWER Flux 5

5

6. Source Range, Neutron Flux 5 10 counts per second 5 1.3 x 10 counts per second i

7. Overtemperature AT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note 3

9. Pressurizer Pressure--Low 1 1945 psig 1 1935 psig kk

10. Pressurizer Pressure--High

-< 2385 psig -< 2395 psig aa

11. Pressurizer Water Level--High 5 92% of instrument span 5 93% of instrument span gg bh

12. Low Reactor Coolant Flow 1 90% of design flow per loop

1 89% of design flow per loop
22 hh
Design flow is 98,400 gpm per loop for Unit I and 95,500 gpm per loop for Unit 2.

TABLE 2.2-1 (Continued) ? REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS m FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E Q c. Power Range Neutron Flux, P-8, 5 48% of RATED 5 49% of RATED Low Reactor Coolant Loop Flow, THERMAL POWER THERMAL POWER H and Reactor Coolant Pump Breaker s Position a d. Low Setpoint Power Range Neutron 10% of RATED > 9%, 5 1EE of RATED Flux, P-10, Enable Block of THERMAL POWER THERMAL POWER Source Intermediate and Power Range Reactor Trips e. Turbine Impulse Chamber Pressure, P-13, Input to Low Power Reactor 5 10% RTP Turbine 5 II% RTP Turbine Trips Block P-7 Impulse Pressure Impulse Pressure 7 Equivalent Equivalent w 19. Reactor Trip Breakers N.A. N.A. 20. Automatic Trip and Interlock Logic N.A. N.A.

TABLE 2.2-1 (Continued) N REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E NOTATION i E ] NOTE 1: OVERTEPPERATURE AT H 1+t S 1 { AT (y,

3) (y, Tag) $ AT, My 2 (1 + t S)U(1 + TcS)-T'] + K ( ~

) ~ # (O )} -K 3 1 m Measured AT by RTD Manifold Instrumentation, Where: AT = 1{ b Lead-lag compensator on measured AT, = 7 3 = Time constants utilized in the lead-lag controller for ti. T2 AT, Ti = 8 sec., tz = 3 sec., m 1 Lag compensator on measured AT, = 3, 13 Time constants utilized in the lag compensator for AT, Ta = 2 sec., = AT, Indicated AT at RATED THERMAL POWER, = K 5 1.0952 (Unit 2),1.4060 (Unit 1), y yy K 0.0133 (Unit 2), 0.0222 (Unit 1), = 2 1+T 5 4 The function generated by the lead-lag controller for T,yg dynamic compensation, = y, 3 zz Time constants utilized in the lead-lag controller for T P.o 14 3 =

avg,

'I 33 sec. (Unit 2), 13 = 4 sec w'M T = 28 sec (Unit 1), 1 = 4 hh Average temperature, F, T = ms 1 Lag compensator on measured T,yg, vv = 3, 3

.i ] TABLE 2.2-1 (Continued) ~ E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

o 7

NOTATION (Continued) E NOTE 1: (Continued) -4 Time constant utilized in the measured T lag compensator, is = 2 sec = To "#9 (Units 1 & 2), 8 T' < 588.1 F Reference T at RATED THERMAL POWER, avg 0.000647 (Unit 2), 0.001095 (Unit 1), K = 3 Pressurizer pressure, psig, P = P' 2235 psig (Nominal RCS operating pressure), = to -1 E S = Laplace transform operator, sec and f (AI) is a function of the indicated difference between top and bottom detectors y of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: (i) for q g between -36% and +8.0% (Unit 2), - 41% and -4.0% (Unit 1); f (AI) = 0, g b y where q and q are percent RATED THERMAL POWER in the top and bottom t b M1 es of 2.g the core respectively, and q

9 is total THERMAL POWER in percent of RATED t

b ! ro THERMAL POWER; 55 (ii) for each percent that the magnitude of q q exceeds -36% (Unit 2), -41% (Unit 1), oo t b the AT Trip Setpoint shall be automatically reduced by 1.173% (Unit 2), 3.151% (Unit 1) $. F of its value at RATED THERMAL POWER; and (iii) for each percent that the magnitude of q q exceeds +8.0% (Unit 2), -4.0% 22 (Unit 1),theATTripSetpointshallbekutomaticallyreducedby0.901% b (Unit 2), 1.447% (Unit 1) of its value at RATED THERMAL POWER. SC

TABLE 2.2-1 (Continued) { REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS S NOTATION (Continued) N$ i NOTE 2: OVERPOWER AT E rS 1 1 Z A T (I * ) ( 1 ) < AT {K -K ( )( ) T -K [T( )- T"] - f (AI)}

I 5

1 + TsS o 4 5 1 + r7S 1+rS 6 1+t S 2 2 c 4 e Where: AT As defined in Note 1, = a b I*'j = As defined in Note 1 = As defined in Note 1 ti,T2 As defined in Note 1, = y, 3 u AT As defined in Note 1, = o o K 5 1.0908 (Unit 2), 1.0708 (Unit 1), 4 0.02/*F for increasing average temperature and 0 for decreasing average K = 5 temperature, T 5 7 1+wS The function generated by the rate-lag controller for T,yg dynamic = 2, > compensation, SS ER T7 Time constant utilized in the rate-lag controller for Tavg, 17 = 5 sec (Units 1 & 2), = ae ee 1 55 As defined in Note 1, = y, Ts3 N$ As defined in Note 1, LL Ts = um 0.00126/ F (Unit 2), 0.00169/ F (Unit 1) for T > T" and 22 K = 6 11 K = 0 for T 5 T", 6 c+ c+ b

TABLE 2.2-1 (Continued) N REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E m NOTATION (Continued) E As defined in Note 1, Q T = m < 588.1 F Reference T at RATED THERMAL POWER, H T" = o, avg o As defined in Note 1, and A-S = to f (AI) 0 for all AI. = 2 Note 3: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2%. l l M 1 c t l Sea l aa FF Uk! L 22 l- = s. t OU l

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 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 from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been "related to DNB through the WRB-1 correlation. The WRB-1 DNB correlation has been developed to pradict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB. The minimum value of the DNBR during steady-state operation, normal ~ operational transients, and anticipated transients is limited to 1.30 (based upon W-3 correlation). This value corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to ONB for all operating conditions. The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid. These curves are based on an enthalpy hot channel factor, F f 1.55 and a reference cosine with a peak of 1.55 f r axial power shape.H,An N j allowance is included for an increase in F at reduced power based on the aH expression: h F = 1.55 [1+ 0.2 (1-P)] H 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 assuming the axial power imbalance is within the limits of the f1 (AI) function of the Overtemperature trip. When the axial power imbalance s 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. McGUIRE - UNITS 1 and 2 B 2-1 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

SAFETY LIMITS BASES ~ For Unit 1, the DNB design basis is as follows: there must be at least a 95% probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR set such that there is a 95% probability with 95% confidence that DNB will not occur when the minimum DNBR is at the DNBR limit. In meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95% confidence that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties. The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor e Coolant System pressure, and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit u ]c is less than the enthalpy of saturated liquid. N The curves are based on a nuclear enthalpy rise hot channel factor, F H, of 1.49 and a reference cosine with a peak of 1.55 for axial power shape. An allow-N ance is included for an increase in F H at reduced power based on the expression: NF H = 1.49 [1 + 0.3 (1-P)] Where P is the fraction of RATED THERMAL POWER. These limiting heat flux conditions are higher than those ca.lculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within tha limits of the f t (AI) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature AT trips will reduce the setpoints to provide protection consistent with core safety limits. 2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radio-nuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements. The entire Reactor Coolant System is hydrotested at 3107 psig,125% of design pressure, to demonstrate integrity prior to initial operation. McGUIRE - UNITS 1 and 2 B 2-2 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

LIMITING SAFETY SYSTEM SETTINGS BASES j Power Range. Neutron Flux (Continued) The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint. Power Range, Neutron Flux, High Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level. Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power. The Power Range Negative Rate trip provides protection for control rod drop accidents. At high power, a rod drop accident of a single or multiple rods could cause local flux peaking which could cause an unconservative local DNBR to exist. The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which DNBR's will be greater than 1.30. 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 uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition. These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux chgnnels. The Source Range channels will initiate a Reactor trip at about 10 s counts per second unless manually blocked 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. I. McGUIRE - UNITS 1 and 2 B 2-4 Amendment No.32 (Unit 1) Amendment NoJ3 (Unit 2) .~

l l j 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL SHUTDOWN MARGIN - T >200 F ag LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% delta k/k (Unit 2),1.3% delta k/k (Unit 1) for four loop operation. APPLICABILITY: MODES 1, 2*, 3, and 4. ACTION: With the SHUTDOWN MARGIN less than 1.6% delta k/k (Unit 2), 1.3% delta k/k (Unit 1), immediately initiate and continue baration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% delta k/k (Unit 2), 1.3% delta k/k (Unit 1): a. Within 1 hour after 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 K,ff greater than or equal to 1.0 at least once per 12 hours by verifying that control bank withdrawal is within the limits of Specification 3.1.3.6; c. When in MODE 2 with K less than 1.0, within 4 hours prior to eff achieving reactor criticality by verifying that the predicted critical control rod position is within the limits of Specification } 3.1.3.6; a d. Prior to initial operation above 5% RATED THERMAL POWER after each i ). fuel loading, by consideration of the factors of Specification l 1 4.1.1.1.le., below, with the control banks at the maximum insertion limit of Specification 3.1.3.6; and

See Special Test Exception 3.10.1.

McGUIRE - UNITS 1 and 2-3/4 1-1 Amendment No.32 (Unit 1) 4 Amendment No.13 (Unit 2)

REACTIVITY CONTROL SYSTEMS [ SURVEILLANCE REQUIREMENTS (Continued) When in MODE 3 or 4, at least once per 24 hours by consideration of e. the following factors: 1) Reactor coolant system boron concentration, 2) Control rod position, 3) Reactor coolant system average temperature, 4) Fuel burnup based on gross thermal energy generation, 5) Xenon concentration, and 6) Samarium concentration. The overall core reactivity balance shall be compared to predicted 4.1.1.1.2 1% delta k/k at least once per values to demonstrate agreement withinThis comparison shall consider at least 31 Effective Full Power Days (EFPD). The predicted those factors statcd in Specification 4.1.1.1.le., above. reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power / Days after each fuel loading. I 4 McGUIRE - UNITS 1 and 2 3/4 1-2

i \\ REACTIVITY CONTROL SYSTEMS ( SHUTDOWN MARGIN - T,yg < 200*F LIMITING CONDITION FOR OPERATION 3.1.1. 2 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% delta k/k. APPLICABILITY: MODE 5. ACTION: With the SHUT 00WN MARGIN less than 1.0% delta k/k, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored. SURVEILLANCE REQUIREMENTS i 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% delta k/k: 4 a. Within 1 hour after 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 i SHUTDOWN MARGIN shall be verified acceptable with an increased j allowance for the withdrawn worth of the immovable or untrippable control rod (s); and l b. At least once per 24 hours by consideration of the following factors: I 1) Reactor Coolant System boron concentration, 2) Control rod position, 3) Reactor Coolant System average temperature, 4) Fuel burnup based on gross thermal energy generation, j 5) Xenon concentration, and 6) Samarium concentration. l 4 i t McGUIRE - UNITS 1 and 2 3/4 1-3

l ? REACTIVITY CONTROL SYSTEMS i l MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION l l i 3.1.1.3 The moderator temperature coefficient (MTC) shall be. i I i j. a. For Unit 1, less positive than the limits shown in Figure 3.1-0, a b. For Unit 2, less positive than 0 delta k/k/*F for the all rods withdrawn, i beginning of cycle life (BOL), hot zero THERMAL POWER condition; and c. For Units 1 and 2, less negative than -4.1 x 10 4 delta k/k/*F for the all rods withdrawn, end of cycle life (EOL), RATED THERMAL POWER condition. APPLICABILITY: Specifications 3.1.1.3a. and 3.1.1.3b. - MODES 1 and 2* only.# Specification 3.1.1.3c. - MODES 1, 2, and 3 only.# ] ACTION: I j a. With the MTC more positive than the limit of Specifications 3.1.1.3a. or 3.1.1.3b, above, operation in MODES 1 and 2 may proceed provided: i j 1. For Unit 1, control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than

l the limits shown in Figure 3.1-0 within 24 hours or be in HOT STANDBY within the next 6 ho'urs. These withdrawal limits shall j

be in addi, tion to the insertion limits of Specification 3.1.3.6; i i 2. For Unit 2, control rod withdrawal limits are established an i maintained sufficient to restore the MTC to less positive than i 0 delta k/k/*F within 24 hours or be in HOT STANOBY within the i next 6 hours. These withdrawal limits shall be in addition to i the insertion limits of Specification 3.1.3.6; i 3. The control rods are maintained within the withdrawal limits j established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and i 4 A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the l I value of the measured MTC, the interim control rod withdrawal i limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition. I b. With the MTC more negative than the limit of Specification 3.1.1.3c. above, be in HOT SHUTDOWN within 12 hours.

With K,77 greater than or equal to 1.0.

See Special Test Exception 3.10.3.

j McGUIRE - UNITS 1 and 2 3/4 1-4 Amendment No.32 (Unit 1) i Amendment No.13 (Unit 2)

,=_ =. _ _ - REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.1.3 The MTC shall be determined to be within its limits during each fuel cycle as follows: a. The MTC shall be measured and compared to the BOL limit of Specifications 3.1.1.3a.and 3.1.1.3b., above, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading; and b. The MTC shall be measured at any THERMAL POWER and compared to -4 -3.2 x 10 delta k/k/*F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm. In the event this compsrison indicates the MTC is more negative than -3.2 x 10 4 delta k/k/*F, the MTC shall be remeasured, and compared to the EOL MTC limit of Specifica-tion 3.1.1.3c., at least once per 14 EFPD during the remainder of the fuel cycle. t McGUIRE - UNITS 1 and 2 3/4 1-5 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

, UNIT 1 ? 0.5 5 g 0.4 a 'G Acceptable Unacceptable E Operation Oceration j 0.3 3 0= 0.2 -TN 5 3 0.1

P 0

10 20 30 40 50 60

70 80 90 100

% of Rated Thermal Power FIGURE 3.1-0 MODERATOR TEMPERATURE COEFFICIENT VS POWER LEVEL (UNIT 1) McGUIRE - UNITS 1 and 2 3/4 1-Sa Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

~.. _ -. _ e REACTIVITY CONTROL SYSTEMS / MINIMUM TEMPERATURE FOR CRITICALITY I LIMITING CONDITION FOR OPERATION .+ I 3.1.1.4 The Reat. tor Coolant System lowest operating loop temperature (T'V9) shall be greater than or equal to 551*F. APPLICABILITY: MODES 1 and 2 ACTION: With a Reactor Coolant System operating loop temperature (T@V8e) less than 551*F, restore T to within its limit within 15 minutes o in HOT STANOBY withinthenexti$9 minutes. 1 SURVEILLANCE REQUIREMENTS i 4.1.1.4 The Reactor Coolant System temperature (Tavg) shall be determined to i be greater than or equal to 551*F: Within 15 minutes prior to achieving reactor criticality, and a. b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T is less than 561*F with the T -T avg ref l DeviationAlarmnotresegg 1 I l l l j

With K greater than or equal to 1.0.

I

SeeSpINalTestException3.10.3.

l i i j ( McGUIRE - UNITS 1 and 2 3/4 1-6 +

-= U j. REACTIVITY CONTROL SYSTEMS i ROD DROP TIME I t. LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length shutdown and control rod drop time from the fully withdrawn position shall be less than or equal to 3.3 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with: T,yg greater than or equal to 551*F, and a. b. All reactor coolant pumps operating. APPLICABILITY: MODES 1 and 2. ACTION: a. With the drop time of any full-length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2. b. With the rod drop times within limits but determined with three reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to (*) of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS i 4.1.3.4 The rod drop time of full-length rods shall be demonstrated through measurement prior to reactor critica? ty: r a. For all rods following each removal of the reactor vessel head, b. For specifically affected individual rods following any maintenance on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and c. At least once per 18 months.

These values left blank pending NRC approval of three loop operation.

McGUIRE - UNITS 1 and 2 3/4 1-19 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

REACTIVITY CONTROL SYSTEMS I SHUTDOWN R0D INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shall be fully withdrawn. APPLICABILITY: MODES la and 2*#. ACTION: With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour either: a. Fully withdraw the rod, or b. Declare the rod to be inoperable and apply Specification 3.1.3.1. SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn: Within 15 minutes prior to withdrawal of any rods in Control a. Banks A, B, C or 0 during an approach to reactor criticality, and b. At least once per 12 hours thereafter. ^See Special Test Exceptions 3.10.2 and 3.10.3.

With Keff greater than or equal to 1.0.

McGUIRE - UNITS 1 and 2 3/4 1-20 m --

REACTIVITY CONTROL SYSTEMS CONTROL ROD INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.6 The control banks shall be limited in physical insertion as shown in Figures 3.1-1 and 3.1-2. APPLICABILITY: MODES 1* and 2*#. ACTION: With the control banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2: a. Restore the control banks to within the limits within 2 hours, or b. Reduce THERMAL POWER within 2 hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the above figures, or c. Be in at least HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours.

See Special Test Exceptions 3.10.2 and 3.10.3.

With K,7f greater than or equal to 1.0.

I McGUIRE - UNITS 1 and 2 3/4 1-21 i

] (Fullywithdrawn) 228 a 220 $(29%,228) ,. ' (79%,228)[ ~~ "v.=~y~_'" BANK B ,/.'. 200 /'-- v ,s 180 _f 5 d(0".,162)

o 160- -=

j T (100%, 1 ii i n= ./; z.' 2 140 ./-.__ BANK C m a. $ 120 ,f.' 4d j e 2 100 '/ m O ,f BANK D n-k ga / ^- g ca o E 60 f(0%,47) / 40 -= 20 f'I / 0 0 20 40 60 80 100 (Fullyinserted) RelativePower(Percent) FIGURE 3.1-1 ROD BANK INSERTION LIMITS VS RELATIVE POWER McGUIRE - UNITS 1 and 2 3/4 1-22 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (UNIT 1) LIMITING CONDITION FOR OPERATION l 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the allowed operational space defined by Figure 3.2-1. APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER *. ACTION: a. With the indicated AFD outside of the Figure 3.2-1 limits, 1. Either restore the indicated AFD to within the Figure 3.2-1 limits within 15 minutes, or 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes 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. b. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the Figure 3.2-1 limits. SURVEILLANCE REQUIREMENTS i 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by: a. Monitoring the indicated AFD for each OPERABLE excore channel: 1. At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2. At least once per hour for the first 24 hours after restoring the AFD Monitoring Alarm to OPERABLE status, b. Monitoring and logging the indicated AFD for each OPERABLE excore i~ channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AFD Monitor Alarm 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 limits when at least two OPERABLE excore channels are indicating the AFD to be outside the limits.

See Special Test Exception 3.10.2.

McGUIRE - UNITS 1 and 2 3/4 2-1 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

POWER DISTRIBUTION LIMITS 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (UNIT 2) LIMITING CONDITION FOR OPERATION ) 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target band (flux difference units) about the target flux difference: a. t 5% for core average accumulated burnup of less than or equal to 3000 MWD /MTU, and b. + 3% -12% for core average accumulated burnup of greater than 3000 MWD /MTV. APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER *. ACTION: a. With the indicated AFD outside of the above required target band about the target flux difference and with THERMAL POWER: 1. Above 90% of RATED THERMAL POWER, within 15 minutes either: a) Restore the indicated AFD to within the target band limits, or b) Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER. 2. Between 50% and 90% of RATED THERMAL POWER: a) POWER OPERATION may continue provided: 1) The indicated AFD has not been outside of the above required target band for more than 1 hour penalty deviation cumulative during the previous 24 hours, and 2) The indicated AFD is within the limits shown on Figure 3.2-1. Otherwise, reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes 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. b) Surveillance testing of the Power Range Neutron Flux channels may be performed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within the limits of Figure 3.2-1. A total of 16 hours operation may be accumulated with the AFD outside of the target band during this testing without penalty deviation, b. THERMAL POWER shall not be increased above 90% of RATED THERMAL POWER unless the indicated AFD is within the above required target band and ACTION a.2.a) 1), above has been satisfied.

See Special Test Exception 3.10.2.

McGUIRE - UNITS 1 and 2 3/4 2-2 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

j. }, f' POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION i l ACTION (Continued) c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD has not been outside of the above required target band for more than 1 hour penalty deviation cumulative during the previous 24 hours. Power increases above 50% of RATED THERMAL POWER do not require being within the target band provided the accumulative penalty deviation is not violated. i l SURVEILLANCE REQUIREMENTS 1 4.2.1.1 The indicated AFD shall be determined to be within its limits during j POWER OPERATION above 15% of RATED THERMAL POWER by: b a. Monitoring the indicated AFD for each OPERABLE excore channel: 1) At least o'nce per 7 days when the AFD Monitor Alarm is OPERABLE, and i 2) At least once per hour for the first 24 hours after restoring i the AFD Monitor Alarm to OPERABLE status. l b. Monitoring and logging the indicated AFD for each OPERABLE excore i channel at least once per hour for the.first 24 hours and at least i once per 30 minutes thereafter, when the AFD Monitor Alarm is j inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging. I i 4.2.1.2 The indicated AFD shall be considered outside of its target band when j two or more OPERABLE excore channels are indicating the AFD to be outside the J target band. Penalty deviation outside of the target band shall be accumulated }, on a time basis of: l 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 j outside of the target band at THERMAL POWER levels between 15% and j. 50% of RATED THERMAL POWER. 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. The [: provisions of Specification 4.0.4 are not applicable. L 4.2.1.4 The target flux difference shall be updated at least once per i, 31 Effective Full Power Days by either determining the target flux difference p pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and 0% at the end of the cycle life. The provi-sions of Specification 4.0.4 are not applicable. McGUIRE - UNITS 1 and 2 3/4 2-3 Amendment No. 32 (Unit 1) i' Amendment No.13 (Unit 2)

-c:: c: Ef: K b 5: ei = w (-15,100) (6,100) 100 UNACCEPTABLEr

UNACCEPTABLE OPERATION OPERATION 80 ACCEPTABLE OPERATION

- [- 60 (-31,50) (17,50) 40 20 _ _ _ _ _ = 0 -50 -40 ,-30 -20 -10 0 10 20 30 40 50 Flux Difference (aI)% FIGURE 3.2-la AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER (UNIT 1) McGUIRE - UNITS 1 and 2 3/4 2-4 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

i iEc: : =w-

3-

@=o-Ho : grJ= <=<. c:_;is

u.ii$ i

Oi*r l i

i# ~ 100 UNACCEPTABLE $( 11,90) ! ! (11,90,)_EUNACCEPTABLE T)PERATION ,g 0,P, ERA.T,10N,, _ / i 80 / \\ 1 i ?'. c'-NCCEPtABLEEOPERhTiON M. ~ 60 ( 31,50). (31,50) 40 \\ 20 0 50 40 -30 20 10 0 10 20 30 40 50 FLUX DIFFERENCE (AI) % FIGURE 3.2-Ib AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER (UNIT 2) McGUIRE - UNITS 1 and 2 3/4 2-5 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

1 POWER'DIST'RIBUTION LIMITS 3/4.2.2 HEATFLUXH0_TCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships: q F (Z) 1 [2.32] [K(Z)] for P > 0.5 (Unit 2) q F (Z) 1 [2 15] [K(Z)] for P > 0.5 (Unit 1) q p F (Z) 1 [2 [K(Z)] for P $ 0.5 (Unit 2) q F (Z) 5 [2 ] [K(Z)] for P $ 0.5 (Unit 1) q Where: P = THERMAL POWER RATED THERMAL POWER

and K(Z) is the function obtaired from Figure 3.2-2 for a given core height location.

APPLICABILITY: MODE 1. ACTION: With F (Z) exceecing its limit: q a. Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit 0 within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION { may proceed fdr up to a total of 72 hours; subsequent POWER OPERATION may p.roceed provided the Overpower Delta T Trip Setpoints i (value of K.) have been reduced at least 1% (in AT span) for each 1% F (Z) exceeds the limit; and 0 b. Identify a.nd correct the cause of the out-of-limit condition prior to increesing THER0 L POWER above the reduced limit required by ACTION a,, above; THERMAL POWER may then be increased provided F (Z) i is demonstrated through incore mapping to be within its limit, q i McGUIRE - UNITS 1 and 2 - 3/4 2-5 Amendment No.32 (Unit 1) Amendment No 13 (Unit 2)

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (UNIT 1) 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 F (z) shall be evaluated to determine if F (z) is within its limit by: q q a. Using the movable incore detectors to obtain a 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 q map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. c. Satisfying the following relationship: M 5 x K(z) for P > 0.5 Fq (z) 1 () ) f r P 1 0.5 N 2 x Fq (z) < Z) x 0 where F (z) is the measured F (z) increased by the allowances for q manufacturing tolerances and measurement uncertainty, 2.15 is the Fq limit, K(z) is given in Figure 3.2-2, P is the relative THERMAL POWER, and W(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation. This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9. M d. Measuring Fq (z) according to the following schedule: 1. Upon achieving equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,* or q 2. At least once per 31 Effective Full Power Days, whichever occurs first.

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 a power distribution map obtained.

McGUIRE - UNITS 1 and 2 3/4 2-7 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (UNIT 1) (Continued) e. With measurements indicating I M (7)I F maximum g (K(z)/ over z M has increased since the previous determination of Fq (z) either of the following actions shall be taken: M q (z) shall be increased by 2% over that specified in Specifi-1) F cation 4.2.2.2c. or M P9we(z)shallbemeasuredatleastonceper7EffectiveFull 2) F r Days until two successive maps indicate that ff (z) is not increasing. maximum (K(z)) over z f. With the relationships specified in Specification 4.2.2.2c. above not being satisfied: 1) Calculate the percent F (z) exceeds its limit by the following q express 1oa: Imaximum M I

U l 1p x 100 for P > 0.5 0

x K(z}j j over z g 2 15 p 'I Imaximum " M(z) x W(z) F0 -1f x 100 for P < 0.5

, b x K(zl J

2) Either of the following actions shall be taken: a) Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits of 3.2-1 ta 3 f,FD for each percent F (z) exceeds its limits q as Ger' >.d in Specification 4.2.2.2f.1). Within ' h,, s eset the AFD alarm setpoints to these modified h,, -l f. b) t.omply with the requirements of Specification 3.2.2 for F (z) exceeding its limit by the percent calculated above. q l McGUIRE - UNITS 1 and 2 3/4 2-8 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2) m

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (UNIT 1) (Continued) g. The limits specified in Specifications 4.2.2.2c, 4.2.2.2e., and 4.2.2.2f. above are not applicable in the following core plane regions: 1. Lower core region from 0 to 15%, inclusive. 2. Upper core region from 85 to 100%, inclusive. ofSpecificatioO(Z)ismeasuredforreasonsotherthanmeetingtherequirements 4.2.2.3 When F 4.2.2.2 an overall measured F (z) shall be obtained from a power q distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. McGUIRE - UNITS 1 and 2 3/4.2-9 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

t 8 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (UNIT 2) 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.2.2 F shall be evaluated to determine if F (Z) is within its limit by: xy q a. Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER, b. Increasing the measured F component of the power distribution map xy by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties, c. Comparing the F computed (F ) obtained in Specification 4.2.2b., xy above, to: 1) The F limits for RATED THERMAL POWER (FRTP) for the appropriate xy x measured core planes given in Specifications 4.2.2.2e. and f., below, and 2) The relationship: L R F =p [1+0.2(1-P)], xy Where F ' is the limit for fractional THERMAL POWER operation RTP express as a function of F and P is the fraction of RATED x THERMAL POWER at which F was measured. xy d. Remeasuring F according to the following schedule: xy P 1) When F is greater than the F limit for the appropriate x measured core plane but less than the F relationship, xy additional power distribution maps shall be taken and F P compared to F and F either: x a) Within 24 hours after exceeding by 20% of RATED THERMAL C POWER or greater, the THERMAL POWER at which F*Y was last determined, or b) At least once per 31 EFPD, whichever occurs first. McGUIRE - UNITS 1 and 2 3/4 2-10 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

l POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (UNIT 2) P 2) When the F is less than or equal to the F limit for the xy appropriate measured core plane, additional power distribution maps shall be taken and F*Y compared to F*RTP and F*Y at least C l Y once per 31 EFPD. e. The F limits for RATED THERMAL POWER (F P) shall be provided for xy all core planes containing Bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specifi-cation 6.9.1.9, f. The F limits of Specification 4.2.2.2e., above, are not applicable xy in the following core planes regions as measured in percent of core height from the bottom of the fuel: 1) Lower core region from 0 to 15%, inclusive, 2) Upper core region from 85 to 100%, inclusive, 3) Grid plane regions at 17.8 + 2%, 32.1 + 2%, 46.4 + 2%, 60.6 +- 2%and74.9+2%, inclusive 7and 4) Core plane regions within + 2% of cre height (+ 2.88 inches) about the bank demand position of the Bank "D" control rods. g. With F exceeding Fx, the effects of F on F (Z) shall be xy q evaluated to determine if F (Z) is within its limits. q 4.2.2.3 When F (Z) is measured for other than F determinations, an overall q xy measured F (Z) shall be obtained from a power distribution map and increased q by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. McGUIRE - UNITS 1 and 2 3/4 2-11 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

UNIT 1 al-E lmis si=s =i.

==is =W =i=ls!E mie is!s =f=s! Els E "iEli=!!!E E.!E-E=iE EMM E=iEi: ii!m42!i!#E ~:lili"i SEidEst =JEs": "iMil 9! E 5lE REJ E i'.!!E - 5!MilEiiiMi!iit:iif -:ili=i # sEl#iM EX & (0.10) ! NIE! Ele =!i=~ =E!=" i-(8.10) lii ' !$li;lisi = iE U"lE =E!Ei-1.0 =u g....g=; =ji;Ej. = ;-- g. . 1-p l..... ._.t=n.n g. jf = t g j-j; p p, . ll?i;lu... e: g E !5li555MI:~-- "'i is : :._., "" i311.096,0.936) 2'}E5l5i55.'5}!E 5}:15 i - - - _l - - - L-..4 =. - u-y=2l =lm n rx._', ug=

n 2,ns 3..._p==x.= =;. - :-

. 4 2 = :.= =2= -.y"___l __-..__. - t - - - :2 :=4 = p==- =-- > ===="..... _=d.d. = =I"

2 = j g._;=

= = _,...... t= -- a = = _. :- ~~ -

= = = :--==

t

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0.2 0 2 4 6 8 10 12 BOTTOM CORE HEIGHT ( FEET ) TOP K(Z) - NORMALIZED F (Z) AS J Tb CORE HEIGHT (UNIT 1) q l McGUIRE - UNITS 1 and 2 3/4 2-12 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2) i

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.}... .[ .g.; g; " :in o e N. O. e. W. o. N. Q W e o o o (zP:t aaznVW80N - (ZDI McGUIRE - UNITS 1 and 2 3/4 2-13 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R, R2 shall be maintained within the region of allowable operation 1 shown on Figure 3.2-3 for four loop operation: Where: N N a. R (Unit 1) = 1.49 [1.0 + 3 (1.0 - P)] 1 'R (Unit 2) = 1.49 [1.0 + 2 (l. fi 7 t R1 2 (Unit 1) = R, R b. R (Unit 2) * [1-RBP(BU)] t 2 THERMAL POWER c* P = RATED THERMAL POWER l H=MeasuredvaluesofFhobtainedbyusingthemovableincore d. F i detectors to obtain a power distribution map. The measured values of F shall be used to calculate R since Figure 3.2-3 H includes penalties for undetected feedwater venturi fouling of 0.1% and for measurement uncertainties of 1.7% for flow and 4% forincoremeasurementofFh,and e. R8P (BU) = Rod Bow Penal'.y as a function of region average burnup as shown in Figure 3.2-4, where a region is defined as those assemblies with the same loading date (reloads) or enrich-ment (first core). (Applies to Unit 2 only). APPLICABILITY: MODE 1. ACTION: I With the combination of RCS total flow rate and R, R2 outside the region of 1 acceptable operation shown on Figure 3.2-3: a. Within 2 hours either: 1. Restore the combination of RCS total flow rate and R, t R to within the above limits, or 2 2. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint 4 to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. McGUIRE - UNITS I and 2 3/4 2-14 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

i POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION: (Continued) b. Within 24 hours of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combination of R, R, and RCS total flow rate are restored to 1 2 within the above limits, 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 the combination of R, R2 and indicated t RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the retion of acceptable operation shown on Figure 3.2-3 prior to exceeding the following THERMAL POWER levels: 1. A nominal 50% of RATED THERMAL POWER, 2. A nominal 75% of RATED THERMA POWER, and 3. Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 The combination of indicated RCS total flow rate determined by process computer readings or digital voltmeter measurement and R, and R2 t shall be within the region of acceptable operation of Figure 3.2.3: 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. 4.2.3.3 The indicated RCS total flow rate shall be verified to be within the region of acceptable operation of Figure 3.2-3 at least once per 12 hours when the most recently obtained values of R t and R, obtained per Specification 2 4.2.3.2, are assumed to exist. 4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. 4.2.3.5 The RCS total flow rate shall be determined by precision heat balance measurement at least once per 18 months. McGUIRE - UNITS 1 and 2 3/4 2-15 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

1 0 50 I UNACCEPTABLE o PENALTIES OF 0.1% FOR UNDETECTED FEED-OPERATION q WATER VENTURI FOULING AND MEASUREMENT REGION m 48 UNCERTAINTIES OF 1.7% FOR FLOW AND 4% FOR INCORE MEASUREMENT OF F ARE (1.04, 47.65) AH E INCLUDED IN THIS FIGURE. A 1 { 46 ~ o a a e N C E A su ~ a N 44 8 v ACCEPTABLE o h OPERATION g REGION Y UNACCEPTABLE H 42 w OPERATION g REGION 8 N = m M ACCEPTABLE OPERAllON REGION FOR $98E RTP (1.0, 39.36) <96% RTP (1.0, 38.97) 394% RTP (1.0, 38.57) yy 38 <;92% RTP gg $90% RTP (1.0, 37.39) E$ aa 33 de ,5,5 0.09 0.92 0.94 0.96 0.98 1.00 1.02 1.04 1.06 1.08 w W, 22 R, = F 11.49[1.0 + 0.3 (1.0-P)] i hh UC FIGURE 3.2-3a RCS TOTAL FLOW RATE VERSUS R (UNIT 1)

f PENALTIES OF 0.1% FOR UNDETECTED FEED-5. 46 WATER VENTURI FOULING AND MEASUREMENT g UNCERTAINTIES OF,1.7% FOR FLOW AND 4% e FOR INCORE MEASUREMENT OF F ARE AH p INCLUDED IN THIS FIGURE. 1 N ACCEPTABLE OPERATION ~ k M P REGION FOR m c R ONLY a a C m us F ACCEPTABLE k OPERATION 3: 42 REGION FOR (1.031, 42.0) o R. Er R, u. k UNACCEPTABLE 1-w} F OPERATION i REGION g O u i ACCEPTABLE OPERATION REGION FOR $98% RTP (1.0, 38.888) 596% RTP (1.0, 38.499) 38 594% RTP ,(1.0, 38.110) j TE 592% RTP (1.0, 37.721) gg 590% RTP (1.0, 37.332) gg (1.0, 37.944) oo @e z= 36 .o P 0.09 0.92 0.94 0.96 0.98 1.00 1.02 1.04 1.06 CM 22 N s,, s ,R = FAH 11.49(1.0 + 0.2 (1.0-P)] ea R = Ril(1 - R8PIBU)] yg i Figure 3.2-3b RCS FLOW RATE VERSUS R and R - FOUR LOOPS IN OPERATION (Unit 2) 3 2 I e

k R M 4 -e g M q .t's s 0- >= E I E m R a \\ 2 u. o= z b z o E 5 E m m S_ o m G 4 e 2 Z O E o e 3 = o" a e ?9 m u 8 "u. 4 4 Q Q o (NOll3VH:1) A11VN3d MOG COW McGuire - Units 1 and 2 3/4 2-18 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

= POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO i 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: l 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 reduced to within its limit, or i 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 1.0 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours. 3. Verify that the QUADRANT POWER TILT RATIO is within its limit with n 24 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip i Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours; and j 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%- q j or greater RATED THERMAL POWER. 4 l l

See Special Test Exception 3.10.2.

t McGUIRE - UNITS 1 and 2 3/4 2-19 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

=_ = POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 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: a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL j POWER. 2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of j 1.0, within 30 minutes; 3. Verify that the QUADRANT POWER TILT RATIO is within its limit l within 2 hours after exceeding the limit or reduce THERMAL l 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 THERMAL POWER within the next 4 hours; and 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 l once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER. i c. With the 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 within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. McGUIRE - UNITS 1 and 2 3/4 2-20 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2) l L.

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 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 POW 50% of RATED THERMAL

POWER, 2.

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0, 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 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 THERMAL POWER within the next 4all be determined to be within the limit above 50% of RATED THERMAL POWER by: e a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE, and b. Calculating the ratio at least once per 12 hours during steady-state operation when the alarm is inoperable. 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 from two sets of four symmetric thimble locations or a full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours. McGUIRE - UNITS 1 and 2 3/4 2-21 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DN6 related parameters shall be maintained within the limits shown on Table 3.2-1: Reactor Coolant System T,yg, and a. b. Pressurizer Pressure. 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. SURVEILLANCE REQUIREMENTS 4.2.5 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours. McGUIRE - UNITS 1 and 2 3/4 2-22 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

f f I TABLE 3.2-1 DNB PARAMETERS r f LIMITS l l Four Loops Three Loops l PARAMETER In Operation In Operation i Reactor Coolant System T,yg < 593'F (**) Pressurizer Pressure > 2230 psai* (**) ] ] i i t " Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of i RATED THERMAL POWER. i 1

- These values left blank pending NRC approval of three loop operation, i

r A 4 i l l 1 l i i i i + i McGUIRE - UNITS 1 and 2 3/4 2-23 Amendment No. 32 (Unit 1) ) Amendment No.13 (Unit 2) ._..e.,., .r,, m,,_m. mr., _ _,, _ .,.,y.,._.,.e.-,..,.....,,

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS COLD LEG INJECTION i LIMITING CONDITION FOR OPERATION 3.5.1.1 Each cold leg injection accumulator shall be OPERA 8LE with: a. The isolation valve open, 1 b. A contained borated water volume of between; i 1) 8022 and 8256 gallons (Unit 1), I 2) 8261 and 8496 gallons (Unit 2), c. A baron concentration of between 1900 and 2100 ppm, d. A nitrogen cover pressure of between 400 and 454 psig, and i e. A water level and pressure channel OPERA 8LE. i APPLICABILITY: MODES 1, 2, and 3*. ACTION: a. With one cold leg injection accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERA 8LE status within 1 hour or be in at least HOT STANOBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. b. With one cold leg injection accumulator inoperable due to the isolation valve being closed, either immediately open the isola-tion valve or be in at least HOT STAN08Y within 1 hour and in HOT SHUTDOWN within the following 12 hours. SURVEILLANCE REQUIREMENTS I 4.5.1.1.1 Each cold leg injection accumulator shall be demonstrated OPERABLE: 1 a. At least once per 12 hours by: )- 1) Verifying the contained barated water volume and nitrogen i cover pressure in the tanks, and 2) Verifying that each cold leg injection accumulator isolation valve is open.

Pressurizer pressure above 1000 psig.

McGUIRE - UNITS 1 and 2 3/4 5-1 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2) i

EMERGENCY CORE COOLING SYSTEMS ( SURVEILLANCE REQUIREMENTS (Continued) b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to EE of tank volume by verifying the boron concentration of the accumulator solution; At least once per 31 days when the RCS pressure is above 2000 psig c. by verifying that power to the isolation valve operator is disconnected by removal of the breaker from the circuit; and d. At least once per 18 months by verifying that each accumulator isolation valve opens automatically under each of the following conditions: 1) When an actual or a simulated RCS pressure signal exceeds the P-11 (Pressurizer Pressure Block of Safety Injection) Setpoint, 2) Upon receipt of a Safety Injection test signal. 4.5.1.1.2 Each cold leg injection accumulator water level and pressure channel shall be demonstrated OPERABLE: At least once per 31 days by the performance of an ANALOG CHANNEL f a. OPERATIONAL TEST, and b. At least once per 18 months by the performance of a CHANNEL CALIBRATION. ( McGUIRE - UNITS 1 and 2 3/4 5-2

I i l EMERGENCY CORE COOLING SYSTEMS I j 3/4.5.4 BORON INJECTION SYSTEM i i BORON INJECTION TANK I i i l I i I. t 4 i ) i i 1 (Deleted] 1 } J 4 i 1 l 1 4 1 i l j i i 1 4 1 i 4 McGUIRE - UNITS 1 and 2 3/4 5-11 Amendment No. 32 (Unit 1) l Amendment No.13 (Unit 2) t

t i

4 EMERGENCY CORE COOLING SYSTEMS f

3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION

3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

a. A contained borated water volume of at least 372,100 gallons, b. A boron concentration of between 2000 and 2100 ppm of boron, c. A minimum solution temperature of 70*F, and d. A maximum solution temperature of 100*F. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: With the RWST inoperable, restore the tank to OPERABLE status within I hour or be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE: a. At least once per 7 days by: 1) Verifying the contained borated water volume in the tank, and 2) Verifying the boron concentration of the water. b. At least once per 24 hours by verifying the RWST temperature when the outside air temperature is either less than 70*F or greater than 100'F. ( McGUIRE - UNITS 1 and 2 3/4 5-12

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 B0 RATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTOOWN MARGIN ensures that: (1) the reactor can be made subcritical from all operating conditions, (2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and (3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,yg. The most restrictive condition occurs at EOL, with T,yg at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncon-trolled RCS cooldown. In the analysis of this accident, a minimum SHUT 00WN MARGIN of 1.6% of delta k/k (Unit 2),1.3% delta k/k (Unit 1) is required to control the reactivity transient. Accordingly, the SHUTOOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T,yg less than 200*F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 1% delta k/k SHUT 00WN MARGIN provides adequate protection. 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT The limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the FSAR accident and transient analyses. The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison. The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MUC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental change in the MOC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This value of the MOC was then transformed into the limiting MTC value -4.1 x 10 4 delta k/k/*F. The MTC value of -3.2 x 10 4 delta k/k/*F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting HTC value of -4.1 x 10 4 k/k/*F. Amendment No. 32 (Unit 1) McGUIRE - UNITS 1 and 2 0 3/4 1-1 Amendment No.13 (Unit 2)

o 3 ll REACTIVITY CONTROL SYSTEMS 1 ( U 8ASES V, 5 MODERATOR TEMPERATURE COEFFICIENT (Continued) The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY I J-This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551'F. This limitation is required to ensure: (1) the moderator temperature coefficient is within it analyzed temperature range. (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature. NOT 3/4.1.2 80 RATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, (5) associated Heat j Tracing Systems, and (6) an emergency power supply from OPERA 8LE diesel ( generators. With the RCS average temperature above 200*F, a minimum of two boron i injection flow paths are required to ensure single functional capability in J the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN l' MARGIN from expected operating conditions of 1.6% delta k/k af ter xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires a' 16,321 gallons of 7000 ppe borated water from the boric acid storage tanks or l 4 75,000 gallons of 2000 ppe borated water from the refueling water storage tank (RWST). With the RCS temperature below 200*F, one 8oron Injection System is acceptable without single failure consideration on the basis of the stable

l reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron

} Injection System becomes inoperable, i ', The limitation for a maximum of one centrifugal charging pump to be OPERA 8LE and the Surveillance Requirement to verify all charging pumps except the required OPERA 8LE pump to be inoperable below 300*F provides assurance j_ i that a mass addition pressure transient can be relieved by the operation of a j single PORV. ( tf McGUIRE - UNITS 1 and 2 8 3/4 1-2

I i j 3/4.2 POWER DISTRIBUTION LIMITS i BASES 1 j P The specifications of this section provide assurance of fuel integrity l during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) maintaining the calculated DNBR in the core at or above the 1 design limit during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical prop- ) erties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria i limit of 2200*F is not exceeded. i j The definitions of certain hot channel and peaking factors as used in j these specifications are as follows: j F (Z) Heat Flux Hot Channel Factor, is defined as the maximum local q j heat flux on the surface of a fuel rod at core elevation Z divided i l by the average fuel rod heat flux, allowing for manufacturing toler-j ances on fuel pellets and rods; l F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of g j the integral of linear power along the rod with the highest integrated j power to the average rod power; and i i F*Y(Z) Radial Peaking Factor, is defined as the ratio of peak power density } to average power density in the horizontal plane at core elevation Z. i R l 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX OIFFERENCE (AFD) assure that the FQN) "PP'" i bound envelope of 2.32 (Unit 2), 2.15 (Unit 1) times the normalized axial j peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes. Target flux difference is determined at equilibrium xenon conditions. The full-length rods may be positioned within the corn in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operation at high power levels. The value of the i target flux difference obtained under these conditions divided by the fraction j of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER d for the associated core burnup conditions. Target flux differences for other 1 THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value j by the appropriate fractional THERMAL POWER level. The periodic updating of t 8 the target flux difference value is necessary to reflect core burnup 1 considerations. i i i l McGUIRE - UNITS 1 and 2 8 3/4 2-1 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2) i i

POWER DISTRIBUTION LIMITS BASES AXIAL FLUX DIFFERENCE (Continued) Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution, suffi-ciently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1 hour penalty deviation limit cumulative during the previous 24 hours is provided for d operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, devia-5 tions of the AFD outside of the target band are less significant. The penalty of 2 hours actual time reflects this reduced significance. Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90% of RATED THERMAL POWER. During operation at THERMAL POWER levels between 50% and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of 1 hour and 2 hours, respectively. Figure B 3/4 2-1 shows a typical monthly target band. For Unit 1, the computer determines the minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if d the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allowed AI-Power operating space and the THERMAL POWER is greater than 50% of RATED THERMAL POWER. E 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure that: (1) the design limits on peak local power density and minimum DNBR are not exceeded, and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS accep-tance criteria limit. Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided: McGUIRE - UNITS 1 and 2 B 3/4 2-2 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

j i i l POWER DISTRIBUTION LIMITS BASES t i HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) a. Contrcl rods in a single group move together with no individual rod insertion differing by more than + 13 steps from the group demand position; b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits. F will be maintained within its limits provided Conditions a. through H

d. above are maintained.

As noted on Figures 3.2-3 and 3.2-4, RCS flow rate N and F may be " traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the measured F is also low) to ensure that the calcu-H lated DNBR will not be below the design DNBR value. lhe relaxation of F as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. R as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts t for F less than or equal to 1.49. This value is used in the various accident H N analyses where F influences parameters other than DNBR, e.g., peak clad tem-perature, and thus is the maximum "as measured" value allowed. R, as defined, 2 allows for the inclusion of a penalty for Rod Bow on DNBR only. Thus, knowing the "as measured" values of Fh and RCS flow allows for " tradeoffs" in excess i of R equal tu 1.0 for the purpose of offsetting the Rod Bow DNBR penalty. Fuel rod bowing reduces the value of DNB ratio. Credit is available to partially offset this reduction. This credit comes from a generic or plant-specific design margin. For McGuire Unit 2, the margin used to partially offset rod bow penalties is 9.1%. This margin breaks down as follows: 1) Design limit DNBR 1.6% 2) Grid spacing K, 2.9% 3) Thermal Diffusion Coefficient 1.2% 4) DNBR Multiplier 1.7% 5) Pitch Reduction 1.7% 1 McGUIRE - UNITS 1 and 2 B 3/4 2-4 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) However, the margin used to partially offset rod bow penalties is 5.9% with the remaining 3.2% used to trade off against measured flow being as much as 2% lower than thermal design flow plus uncertainties. The penalties applied to F to account for rod bow (Figure 3.2-4) as a function of burnup are g consistent with those described in Mr. John F. Stolz's (NRC) letter to T. M. Anderson (Westinghouse) dated April 5, 1979 with the difference being due to the amount of margin each unit uses to partially offset rod bow penalties. For McGuire Unit 1, margin between the safety analysis limit DNBRs (1.47 and 1.49 for thimble and typical cells, respectively) and the design limit DNBRs (1.32 and 1.34 for thimble and typical cells, respectively) is maintained. A fraction of this margin is utilized to accommodate the transition core DNBR penalty (2%) and the appropriate fuel rod bow DNBR penalty (WCAP - 8691, Rev. 1) When an F measurement is taken, an allowance for both experimental error q and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance. When RCS flow rate and F are measured, no additional allowances are necessary prior to comparison with the limits of Figures 3.2-3 and 3.2-4. Measurement errors of 1.7% for RCS total flow rate and 4% for F have been g allowed for in determination of the design DNBR value. The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators. Potential fouling of the feedwater venturi which might not be detected could bias the result from the precision heat balance in a non-conservative manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is included in Figure 3.2-3. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling. The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the accept-able region of operation shown on Figure 3.2-3. McGUIRE - UNITS 1 and 2 8 3/4 2-5 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2)

1 POWER DISTRIBUTION LIMITS 4 5 BASES HEAT FLilX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTQ,R, (Continued) The hot channel factor F (z) is measured periodically and increased by q j a cycle end height dependent power factor, W(Z), to provide assurance that the limit oc the hot channel factor, F (2), it met. W(z) accounts for the effects q of normal cperation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core. The W(z) function for normal operation is provided in the Peaking Factor Limit Report per Specification 6.9.1.9. l 3/4.2.4 QpAURANT POWER TILT RATIO i The QUADRANT POWER TILT RATIO limit assures that the radial power distri- + bution satisfies the design values used in the power capability analysis. Radial cover distribution measurements are made during STARTUP testing and periodically during power operation. The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow ~ identification and correc-tion nf a dropped or minaligned rod. In the evrent such action *does not cor-rect the tilt, the margin for uncertainty on F is reinstated by reducing q the power by 3% from RATED THERMAL POWER for ea:h percent of tilt in excess ) o f 1. 0. For purposes of ' monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT 1 F0WER TILT RATIO.- The incore detector monitoring is done with a full incore flux map or two sats of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. There locations are 1 j C-8, E-5, E-11, H-3, H-13, L-5, L-11, N-8. l 3 /4. 2. r> ONB PARAMETE15 The limits on the DNB-related parameters assure that each of the para-meters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent witt, the initial FSAR assumptions and have been analytically demonstrated i I adequate to maintain a design limit DWOR throughout each analyzed transient. l The 12-hour periodic surveillance of those parameters through instrument readout is suff4cient to ensure that the parameters are restored within their limitt,following load changes and other expected transient operation. 4 f McGUIRE - UNITS 1 and 2 3 3/4 2-6 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2) ---s. .,r e .m, .m--- ~, -. -. -,g--e-.-- 9.- .--,, - + - - - - -. - - - --r------ r--

i 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the design limit during all normal operations and antici-pated transients. In MODES I and 2 with one reactor coolant loop not in oper-ation this specification requires that the plant be in at least HOT STANDBY within 1 hour. In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE. In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor ccolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE. In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE. The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during baron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control. The restrictions on starting a reactor coolant pump with one or more RCS cold legs less than or equal to 300'F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either: (1) restricting the water volume in the pressurizer and thereby providing a volume for the reactor coolant to expand into, or (2) by restricting starting of the RCPs to when the secondary water tempera-ture of each steam generator is less than 50*F above each of the RCS cold leg temperatures. McGUIRE - UNITS 1 and 2 8 3/4 4-1 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

REACTOR COOLANT SYSTEM ( BASES 3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RC5, provides overpressure relief capability and will prevent dCS overpres-surization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures. During operation, all pressurizer Code safety valves must be OPERA 8LE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss of load assuming no Reactor trip until the first Reactor Trip System Setpoint is reached (i.e., no credit is taken for a direct Reactor trip on the loss of load) and also assuming no operation of the power-operated relief valves or steam dump valves. Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI ( of the ASME Boiler and Pressure Code. 3/4.4.3 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR. The limit is consistent with the initial SAR assumptions. The 12 hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation. s McGUIRE - UNITS 1 and 2 B 3/4 4-2 i

(* 3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES 3/4.5.1 ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures. The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met. The accumulator power operated isolation valves are considered to be " operating bypasses" in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required. The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occufring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a made where this capability is not required. 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long-term core cooling capability in the recirculation mode during the accident recovery period. With the RCS temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements. McGUIRE - UNITS 1 and 2 B 3/4 5-1

EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued) The limitation for a maximum of one centrifugal charging pump and one Safety Injection pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and Safety Injection pumps except the required OPERABLE charging pump to be inoperable below 300 F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance Requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout. conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. 3/4.5.4 [ Deleted] 3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWST minimum volume and boron concentration ensure that: (1) sufficient water is available within containment to permit recirculation cooling flow to the core, and (2) the reactor will remain subtritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses. Amendment No. 32 (Unit 1) McGUIRE - UNITS 1 and 2 B 3/4 5-2 Amendment No.13 (Unit 2)

l ADMINISTRATIVE CONTROLS ACTIVITIES (Continued) 6.5.1.6 ALL REPORTABLE EVENTS and all violations of Technical Specifications shall be investigated and a report prepared which evaluates the occurrence and which provides recommendations to prevent recurrence. Such reports shall be approved by the Station Manager and transmitted to the Vice President, Nuclear Production, and to the Director of the Nuclear Safety Review Board. 6.5.1.7 The Station Manager shall assure the performance of special reviews and investigations, and the preparation and submittal of reports thereon, as requested by the Vice President, Nuclear Production. 6.5.1.8 The station security program, and implementing procedures, shall be reviewed at least once per !2 months. Recommended changes shall be approved by the Station Manager and transmitted to the Vice President, Nuclear Produc-tion, and to the Director of the Nuclear Safety Review Board. 6.5.1.9 The station emergency plan, and implementing procedures, shall be reviewed at least once per !2 months. Recommended changes shall be approved by the Station Manager and transmitted to the Vice President, Nuclear Produc-tion, and to the Director of the Nuclear Safety Review Board. 6.5.1.10 The Station Manager shall assure the performance of a review by a qualifisd individual / organization of every unplanned onsite release of radio-active material to the environs including the preparation and forwarding of reports covering evaluation, recommendations, and disposition of the corrective ACTION to prevent recurrence to the Vice President, Nuclear Production and to the Nuclear Safety Review Board. 6.5.1.11 The Station Manager shall assure the performance of a review by a qualified individual / organization of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and Radwaste Treatment Systems. 6.5.1.12 Reports documenting each of the activities performed under Specifi-cations 6.5.1.1 through 6.5.1.11 shall be maintained. Copies shall be provided to the Vice President, Nuclear Production, and the Nuclear Safety Review Board. 6.5.2 NUCLEAR SAFETY REVIEW BOARD (NSRB) FUNCTION 6.5.2.1 The NSRB shall function to provide independent review and audit of designated activities in the areas of: a. Nuclear power plant operations, b. Nuclear engineering, c. Chemistry and radiochemistry, Amendment No. 32 (Unit 1) McGUIRE - UNITS 1 and 2 6-9 Amendment No.13 (Unit 2) ( A 5 w

__.1_----------- -m l ADMINISTRATIVE CONTROLS FUNCTION (Continued) d. Metallurgy, l e. Instrumentation and control, f. Radiological safety, g. Mechanical and electrical engineering, and h. Administrative control and quality assurance practices. ORGANIZATION 6.5.2.2 The Director, members and alternate members of the NSRB shall be appointed in writing by the Vice President, Nuclear Production, and shall have an academic degree in an engineering or physical science field; and in addition, shall have a minimum of 5 years technical experience, of which a minimum of 3 years shall be in one or more areas given in Specifica-( tion 6.5.2.1. No more than two alternates shall participate as voting members in NSRB activities at any one time. 6.5.2.3 The NSRB shall be composed of at least five members, including the Director. Members of the NSRB may be from the Nuclear Production Department, ( from other departments within the Company, or from external to the Company. A maximum of one member of the NSRB may be from the McGuire, Nuclear Station staff. 6.5.2.4 Consultants shall be utilized as determined by the NSRB Director to provide expert advice to the NSRB. 6.5.2.5 Staff assistance may be provided to the NSRB in order to promote the proper, timely, and expeditious performance of its functions. 6.5.2.6 The NSRB shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per 6 months thereafter. 6.5.2.7 The quorum of the NSRB necessary for the performance of the NSRB review and audit functions of these Technical Specifications shall consist of the Director, or his designated alternate, and at least four other NSRB members including alternates. No more than a minority of the quorum shall have line responsibility for operation of McGuire Nuclear Station. i ( McGUIRE - UNITS 1 and 2 6-10 es 3

ADMINISTRATIVE CONTROLS i REVIEW 6.5.2.8 The NSRB shall review: a. The safety evaluations for: (1) changes to procedures, equipment, or systems, and (2) tests or experiments completed under the provision of Section 50.59, 10 CFR to verify that such actions did not constitute an unreviewed safety question; b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR; c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR; d. Proposed changes in Technical Specifications or this Operating License; e. Violations of Codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance; f. Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety; g. All REPORTABLE EVENTS; h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems or components that could affect nuclear safety; i. Quality Assurance Department audits relating to station operations and actions taken in response to these audits; and j. Reports of activities performed under the provisions of Specifi-cations 6.5.1.1 through 6.5.1.11. AUDITS 6.5.2.9 Audits of unit activities shall be performed under the cognizance of the NSRB. These audits shall encompass: a. The conformance of unit operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 months; b. The performance, training, and qualifications of the entire unit staff at least once per 12 months; 9 AmendmentNo.32(Unit 1) McGUIRE - UNITS 1 and 2 6-11 Amendment No.13 (Unit 2)

.. - ~. . _.L. ^ : _ ~~ ADMINISTRATIVE CONTROLS i AUDITS (Continued) c. The results of actions taken to correct deficiencies occurring in unit equipment, structures, systems, or method of operation that affect nuclear safety at least once per 6 months; d. The performance of activities' required by the Operational Quality Assurance Program to meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months; The Emergency Plan and implementing procedures at least once per e. 12 months; f. The Security Plan and implementing procedures at least once per 12 months; g. The Facility Fire Protection programmatic controls including the implementing procedures at least once per 24 months by qualified licensee QA personnel; h. The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant. An outside independent fire protection consultant shall be used at least every third year; i 1. The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months; j. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months; k. The PROCESS CONTROL PROGRAM and implementing procedures for SOLIDIFICATION of radioactive wastes at least once per 24 months; 1. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per 12 months; and Any other area of unit operation considered appropriate by the NSRB m. or the Vice President Nuclear Production. i AUTHORITY t 6.5.2.10 The NSRB shall report to and advise the Vice President, Nuclear Production, on those areas of responsibility specified in Specifications 6.5.2.8 and 6.5.2.9. -i h k McGUIRE - UNITS 1 and 2 6-12

ADMINISTRATIVE CONTROLS RECORDS 6.5.2.11 Records of NSRB activities shall be prepared, approved, and distributed as indicated below: a. Minutes of each NSRB meeting shall be prepared, approved, and forwarded to the Vice President, Nuclear Production, and to the Executive Vice President, Power Operations, within 14 days following each meeting; b. Reports of reviews encompassed by Specification 6.5.2.8 above, shall be prepared, approved and forwarded to the Vice President, Nuclear Production, and to the Executive Vice Presioent, Power Operations, within 14 days following completion of the review; and c. Audit reports encompassed by Specification 6.5.2.9 above, shall be forwarded to the Vice President, Steam Production, and to the Executive Vice President, Power Operations, and to the management positions responsible for the areas audited within 30 days after completion of the audit by the auditing organization. 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS: a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and b. Each REPORTABLE EVENT shall be reviewed by the Station Manager; or by: (1) the Operating Superintendent, (2) the Technical Services Superintendent, or (3) the Maintenance Superintendent, as previously designated by the Station Manager, and the results of the review shall be submitted to the NSRB and the Vice President, Nuclear Production.

6. 7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The NRC Operations Cente? shall be notified by telephone as soon as possible and in all cases within 1 hour. The Vice President, Nuclear Production, and the NSRB shall be notified within 24 hours; b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the Operating Superintendent and the Station Manager. This report shall describe: (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components, systems, or structures, and (3) corrective action taken to prevent recurrence; AmendmentNo.32(Unit 1) Amendment No.13 (Unit 2) McGUIRE - UNITS 1 and 2 6-13

ADMINISTRATIVE CONTROLS ( SAFETY LIMIT VIOLATION (Continued) The Safety Limit Violation Report shall be submitted to the c. Commission, the NSRB and the Vice President, Nuclear Production, within 14 days of the violation; and d. Critical operation of the unit shall not be resumed until authorized by the Commission. 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below: The applicable procedures recommended in Appendix A of Regulatory a. Guide 1.33, Revision 2, February 1978; b. The applicable procedures required to implement the requirements of NUREG-0737; c. Security Plan implementation; d. Emergency Plan implementation; e. PROCESS CONTROL PROGRAM implementation; f. OFFSITE DOSE CALCULATION MANUAL implementation; and g. Quality Assurance Program for effluent and environmental monitoring. 6.8.2 Each procedure of Specification 6.8.1 above, and changes thereto, shall be reviewed and approved by the Station Manager; or by: (1) the Operating Superintendent, (2) the Technical Services Superintendent, or (3) the Maintenance Superintendent, as previously designated by the Station Manager; prior to implementation and shall be reviewed periodically as set forth in administrative procedures. 6.8.3 Temporary changes to procedures of Specification 6.8.1 above may be made provided: a. The intent of the original procedure is not altered; b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator license on the unit affected; and ( McGUIRE - UNITS.1 and 2 6-14

2 . =... ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) c. The change is documented, reviewed, and approved by the Station Manager; or by: (1) the Operating Superintendent, (2) the Technical Services Superintendent, or (3) the Maintenance Superintendent, as previously designated by the Station Manager, within 14 days of implementation. 6.8.4 The following programs shall be established, implemented, and maintained: a. Reactor Coolant Sources Outside Containment 1 A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The i systems include RHR, Boron Recycle, Refueling Water, Liquid Waste, Waste Gas, Safety Injection, Chemical and Volume Control, Contain-ment Spray, and Nuclear Sampling. The program shall include the following: 1) Preventive maintenance and periodic visu'al inspection l requirements, and j 2) Inte, grated leak test requirements for each system at refueling cycle intervals or less. e b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following: 1) Training of personnel, 2) Procedures for monitoring, and 3) Provisions for maintenance of sampling and analysis equipment. c. Secondary Water Chemistry A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shall include: 1) Identification of a sampling schedule for the critical variables and control points for these variables, 2) Identification of the procedures used to measure the values of the critical variables, 3) Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage, McGUIRE - UNITS 1 and 2 6-15

l l l ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) 4) Procedures for the recording and management of data, 5) Procedures defining corrective actions for all off-control point chemistry conditions, and 6) A procedure identifying: (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action. I d. Backup Method for Determining Subcooling Margin A program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include the following: 1) Training of personnel, and 2) Procedures for monitoring. e. Post-accident Sampling A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines, and particulates in plant gaseous effluents, and containment atmosphere samples under accident

}

conditions. The program shall include the following: I 1) Training of personnel, { 2) Procedures for sampling and analysis, and ) 3) Provisions for maintenance of sampling and analysis equipment. 1 6.9 REPORTING REQUIREMENTS l ROUTINE REPORTS i 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the NRC Regional Office unless otherwise noted. L STARTUP REPORT

6. 9.1.1 A summary report of plant STARTUP and power escalation testing shall be submitted following:

(1) receipt of an Operating License, (2) amendment to the License involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. AmendmentNo.32(Unit 1) Amendment No. 13 (Unit 2) McGUIRE - UNITS 1 and 2 6-16

ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) L~ .-o The Radioactive Effluent Releasa Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radio-active Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof. The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability.* This same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figures 5.1-3 and 5.1-4) during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM). The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources, including doses from primary effluent pathways and direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190, " Environmental Radiation Protection Standards for Nuclear Power Operation." Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.

In lieu of submission with the first half year Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

McGUIRE - UNITS 1 and i 6-19

ADMINISTRATIVE CONTROLS SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61)* shipped offsite during the report period: a. Container volume, b. Total Curie quantity (specify whether determined by measurement or estimate), c. Principal radionuclides (specify whether determined by measurement or estimate), d. Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms), e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and f. Solidification agent or absorbent (e.g., cement, urea formaldehyde). The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period. The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2. MONTHLY OPERATING REPORTS 6.9.1.8 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Manage-ment, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the NRC Regional Office, no later than the 15th of each month following the calendar month covered by the report.

These requirements shall not become effective for reporting solid waste shipped offsite before January 1, 1984 and which is to be reported in the July 1984 Semiannual Radioactive Effluents Release Report.

McGUIRE - UNITS 1 and 2 6-20 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2)

11 ~ ~~ i l: ADMINISTRATIVE CONTROLS l RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.9 The F limit for RATED THERMAL POWER (FRTP) shall be provided to xy x the Regional Administrator of the NRC Regional Office, with a copy to the Director, Nuclear Reactor Regulation, Attention: Chief, Core Performance N g Branch, U. S. Nuclear Regulatory Commission, Washington, D.C. 20555 for all z core planes containing Bank "D" control rods and all unrodded core planes at g least 60 days prior to cycle initial criticality. In the event that the limit would be submitted at some other time during core life, it shall be submitted c 60 days prior to the date the limit would become effective unless otherwise exempted by the Commission. RTP Any information needed to support F will be by request from the NRC and x ,need not be included in this report. 'The W(z) function for normal operation shall be provided to the Director, Nuclear Reactor Regulations, Attention: Chief, Core Performance Branch, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 at least 60 days prior to cycle initial criticality. In the event that these values would be submitted at some other time during core life, it will be submitted 60 days

E prior to the date the values would become effective unless otherwise exempted 3

by the Commission. Any information needed to support W(z) will be by request from the NRC and need not be included in this report. SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report. Amendment No. 32 (Unit 1) McGUIRE - UNITS 1 and 2 6-21 AmendmentNo.13(Unit 2)

ADMINISTRATIVE CONTROLS 6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. 6.10.1 The following records shall be retained for at least 5 years: a. Records and logs of unit operation covering time interval at each power level; b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety; c. All REPORTABLE EVENTS; d. Records of surveillance activities; inspections and calibrations required by these Technical Specifications; e. Records of changes made to the procedures required by Specifi-cation 6.8.1; f. Records of radioactive shipments;

g.

Records of sealed source and fission detector leak tests and results; and h. Records of annual physical inventory of all sealed source material of record. 6.10.2 The following records shall be retained for the duration of the unit Operating License: a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report; b. Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories; c. Records of radiation exposure for all individuals entering radiation control areas;. d. Records of gaseous and liquid radioactive material released to the environs; e. Records of transient or operational cycles for those unit components identified in Table 5.7-1; f. Records of reactor tests and experiments; McGUIRE - UNITS 1 and 2 6-22 Amendment No.13 (Unit 2) Amendment No.32 (Unit 1)

ADMINISTRATIVE CONTROLS RECORD RETENTION (Continued) g. Records of training and qualification for current members of the unit staff; h. Records of inservice inspections performed pursuant to these Technical Specifications; i. Records of quality assurance activities required by the QA Manual; j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59; k. Records of meetings of the NSRB and reports required by Specification 6.5.1.12; 1. Records of the service lives of all snubbers listed in Tables 3.7-4a and 3.7-4b including the date at which the service life commences and associated installation and maintenance records; m. Records of secondary water sampling and water quality; and n. Records of analyses required by the Radiological Environmental Monitoring Program that would permit evaluation of the accuracy of the analysis at a later date. This should include procedures effective at*specified times and QA records showing that these procedures were followed. 6.11 RADIATION PROTECfION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)*. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radia-tion areas.

McGUIRE - UNITS 1 and 2 6-23 Amendment No.13 (Unit 2) Amendment No.32 (Unit 1)

-.-.:..~ L ADMINISTRATIVE CONTROLS 4 HIGH RADIATION AREA (Continued) i a. A radiation monitoring device which continuously indicates the radiation dose rate in the area; or 4 b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be.nade after the dose rate level in the area has been established and personnel have been made knowledgeable of them; or c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device l who is responsible for providing positive control over the activities 1-within the area and shall perform periodic radiation surveillance at l the frequency specified by the Station Health Physicist in the RWP. l 6.12.2 In addition to the requirements of Specification 6.12.1, areas i accessible to personnel with radiation levels such that a major portion of the body could receive in 1 hour a dose greater than 1000 mrem shall be provided i with locked doors to prevent unauthorized entry, and the keys shall be main-tained under the administrative control of the Shift Foreman on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for indi-viduals in that area. For individual areas accessible to personnel with radia-tion levels such that a major portion of the body could receive in 1 hour a dose in excess of 1000 mrem

that are located within large areas, such as PWR containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall 4

be roped off, conspicuously posted, and a flashing light shall be activated as a warning device. In lieu of the stay time specification of the RWP, direct or-remote (such as use of closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area. i 6.13 PROCESS CONTROL PROGRAM (PCP) 6.13.1 The PCP shall be approved by the Commission prior to implementation. l' 6.13.2 Licensee-initiated changes,to the PCP: a. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made. This submittal shall contain: I

Measurement made at 18 inches from source of radioactivity.

McGUIRE - UNITS 1 and 2 6-24 Amendment No.32 (Unit 1) Amendment No.13 (Unit 2) i f .~

ADMINISTRATIVE CONTROLS PROCESS CONTROL PROGRAM (PCP) (Continued) 1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information; 2) A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and 3) Documentation of the fact that the change has been reviewed and found acceptable by the Station Manager. b. Shall become effective upon review and acceptance by a qualified individual / organization. 6.14 0FFSITE DOSE CALCULATION MANUAL (ODCM) 6.14.1 The ODCM shall be approved by the Commission prior to implementation. 6.14.2 Licensee-initiated changes to the ODCM: a. ShalT be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective. This submittal shall contain: 1) Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered, dated and containing the revision number together with appropriate analyses or evaluations justifying the change (s); 2) A determination that the change will not reduce the accuracy or reliability of dose calculations or Setpoint determinations; and 3) Documentation of the fact that the change has been reviewed and found acceptable by the Station Manager. b. Shall become effective upon review and acceptance by a qualified individual / organization. l l l l McGUIRE - UNITS 1 and 2 6-25 Amendment No.13 (Unit 2) l Amendment No. 32 (Unit 1) t

_y i 1 ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO RADI0 ACTIVE LIQUID, GASEOUS, AND SOLID WASTE TREATMENT SYSTEMS

6.15.1 Licensee-initiated major changes to the Radioactive Waste Systems (liquid, gaseous, and solid):

Shall be reported to the Commission in the Semiannual Radioactive a. Effluent Release Report for the period in which the evaluation was reviewed by the Station Manager. The discussion of each change shall contain: 1) A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR Part 50.59; 2) Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; 3) A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems; 4) An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the License application and amendments thereto; 5) An evaluation of the change, which shows the expected maximum exposures to individual in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the License application and amendments thereto; 6) A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made; 7) An estimate of the exposure to plant operating personnel as a result of the change; and 8) Documentation of the fact that the change was reviewed and found acceptable by the Station Manager. b. Shall become effective upon review and acceptance by a qualified individual / organization.

Licensees may chose to submit the information called for in this specifi-cation as part of the annual FSAR update.

McGUIRE - UNITS 1 and 2 6-26 Amendment No. 32 (Unit 1) Amendment No.13 (Unit 2) -,.}}

v t e Letter Sequence Other
TAC:53319, (Open) TAC:53320, (Open)
Initiation Request, Request, Request Acceptance... Administration Questions Answers Results Approval Other: ML20082D639, ML20082D657, ML20084C989, ML20086P477
MONTHYEAR1984-11-16 [Table View]

ML20084C989

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_. i =_,r-3_2 _-f.,;.- ;'2_1 m ___c,. _.. -l====l m.l:. _::===:====:t-]_..- i