Text

Amends 86 & 80 to Licenses NPF-35 & NPF-52,respectively, Revising Tech Specs to Reflect Fuel Reloading for Cycle 6 Operation W/Fuel Mfg by B&W Fuel Co
	ML20073R450
Person / Time
Site:	Catawba
Issue date:	05/31/1991
From:	Matthews D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20073R451	List: ML20073R450; ML20073R452;
References
	NUDOCS 9106070203
	Download: ML20073R450 (78)
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g p DUKE POWER COMPANY NORTH CAROLINA FLECTRIC MEttBERSHIP COPPORATION SALUDA RIVER ELECTRIC C00 PERT.TIVE, it!C. 00CKET NO. 50-413 CATAWBA NUCLEAR STAT 10t', UNIT 1 Al'ENDMENT TO FtCILITY OPERATING LICENSE Amendment No. 86 License No. NPF-35 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility) Facility Operating License No. NPF-35 filed by the Duke Power Company, acting for itself, North Carolina Electric Membership) Corporation and Saluda River Electric Cooperative, Inc. (licensees dated Janaury 9,1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; C. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this emendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 4 9106070203 910531 PDR ADOCK 05000413 P ppy i

' ' p e nso / 'o, / "- 'n UNITED STATES { [ . g),i NUCLEAR REGULATORY COMMISSION J o, g WASHINGTON. D C. 20555 DUKE POWER COMPANY ll0RTH CAROLINA MUtilCIPAL POWER AGENCY N0. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCVET NO. 50-414 "~ CATAWBA NUCLEAR STATIO!O, UNIT 2, AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 80 License No. NPF-52 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Facility Operating License No. NPF-5? filed by the Duke Power Company, actino for itself, North Carolina Municipal Power Agency No. 1 and Piedmont Municipal Power Agency (licensees) dated January 9, 1991, complies with the standards and requirements of the Atomic Enercy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regu htions of the Commission; C. .There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. s w

_. =. -. 2. Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendinent, and Paragraph 2.C.(2) of Facility Operating License No. NPF-52 is hereby -amended to read as:follows: Technical Specifications The Technical Specifications contained in Appendix A, as revis.ed through Amendment No. 80 and the Environmental Protection Plan contained-in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Duke Power Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This license amendment is effective as of its date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION / / ( p a rc i David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation

Attachment:

~ Technical Specification Changes Date of Issuance: May 31, 1991 I. l l 9 I w -~~e m w

4. ATTACHMENT TO LICENSE AMEilDMENT NO. 86 FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND TO LICEtlSE AtiENDMEtlT NO. 80 FACILITY OPERATING LICENSE NO. NPF-52 DOCKET 110. 50-414 Replace the following paces of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change. The corresponding over-leaf pages are also provided to maintain document completeness. Remove Pages Insert Pages III III IV IV V Va Yb VII VII* VIII VIII XIII XIIIa XIIIb 2-1 2-1 2-2 2-2 2-2a 2-4 2-4 2-7 2-7 2-8 2-8 2-10 2-10 B 2-1 through B 2-2 B 2-1 through B 2-2a 3/4 2-1 through 3/4 2-16 3/4 A2-1 through 3/4 A2-16 and 3/4 B2-1 through 3/4 B2-16 3/4 5-1 thru 3/4 5-4a 3/4 5-1 thru 3/4 5-4 3/4 5-7 3/4 5-7* 3/4 5-8 3/4 5-8

l B 3/4 2-1 thru 8 3/4 2-6 B 3/4 2-1 thru 8 3/4 2-10 6-19 6-19 6-19a 6-19a

0verleaf pages 1

\\ i SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR C0RE............... 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.......................... 2-1 FIGURE 2.1-la REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION., 2-2 FIGURE 2.1-lb REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.. 2-2a

2. 2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETP0lNTS...............

2-3 TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS.... 2-4 BASES SECTION 2.1 SAFETY LIMI_TS 2.1.1 REACTOR CORE (FOR UNIT 1).................................... B 2-1 2.1.1 REACTOR CORE (FOR UNIT 2).................................... B 2-2 2.1.2 REACTOR COOLANT SYSTEM PRESSURE.............................. B 2-2a 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS.............. B 2-3 4 CATAWBA - UNITS 1 & 2 III Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

a 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 Shutdown Margin - T > 200 F........................... 3/4 1-1 avg Shutdown Margin - T < 200*F.............. 3/4 1-3 avg Moderator Temperature Coefficient...... 3/4 1-4 Minimum Temperature for Cri ticali ty...................... 3/4 1-6 3/4.1.2 B0 RATION 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 Bcrated Water Sources - Operating................... 3/4 1-12 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Grcup Height........................................ 3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE E'.'ENT 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 Orop Time............................................ 3/4 1-19 Shutdown Rod Insertion Limit............................. 3/4 1-20 Control Bank Insertion Limits............................ 3/4 1-21 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (Unit 1)........................... 3/4 A2-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (X,Y,Z) (Unit 1)........ 3/4 A2-3 q 1 CATAWBA - UNITS 1 & 2 IV Amendment No. 86(Unit 1) Amendment No. 80(Unit 2) i I

a e LIMITIhG CONDITIONS FOR OPERATION AND' SURVEILLANCE-REQUIREMENTS SECTION-PAGE 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Unit1)........ 3/4 A2-7 3/4.2.4 QUADRANT POWER TILT RATIO (Unit 1)....................... 3/4 A2-10 3/4.2.5 DNB PARAMETERS (Unit 1).................................. 3/4 A2-13 TABLE 3.2-1 DNB PARAMETERS (Unit 1)............................... 374 A2-15 FIGURE 3.2-1 REACTOR COOLANT SYSTEM TOTAL FLOW RATE VERSUS RATED THERMAL POWER-FoVR LOOPS IN OPERATION (Unit1)........... 3/4 A2-16 3/4.2.1 AXIAL FLUX DIFFERENCE (Unit 2)........................... 3/4 B2-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - FQ(Z) (Unit 2)............ 3/4 B2-3 3/4.2.3 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Unit 2)......................... 3/4 B2-9 3/4.2.4 - QUADRANT POWER TILT RATIO (Unit 2)..... 3/4 B2-12 -3/4.2.5 DNB-PARAMETERS (Unit 2).............. 3/4 82-15 TAB L E ---3. 2-1 DNB PARAMETERS (Unit 2)............................... 3/4 B2-16 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION...................... 3/4 3-1 TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION................... 3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES.... 3/4 3-7 -TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................. 3/4 3-9 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4'3-13 -TABLE 3.3-3. ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................... 3/4 3-15 TABLE 3.3-4 ENGINEERED SAFETY FEATURES AClUATION SYSTEM INSTRUMENTATION TRIP SETP0INTS........................... 3/4 3-27 TABLE 3.3 ENGINEERED SAFETY FEATURES RESPONSE TIMES............. 3/4 3-37 TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM-INSTRUMENTATION SURVEILLANCE REQUIREMENTS................ 3/4 3-42 '4 CATAWBA - UNITS 1 & 2 Va Amendment No.86 (Unit 1) Amendment No.80 (Unit 2)

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION _PAGE 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring For Plant Operations........ 3/4 3-51 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR P LANT 0PERATIONS.................................... 3/4 3-52 l TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS.................... 3/4 3-54 Movable Incore Detectors.............................. 3/4 3-55 Seismic Instrumentation............................. 3/4 3-56 TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTAT REQUIREMENTS...................... ION............ 3/4 3-57 3/4 3-58 Meteorological Instrumentation...................... 3/4 3-59 t l { l l l CATAWBA - UNITS 1 & 2 Vb Amendment No. 86 (Unit 1) l Amendment No. 80 (Unit 2)

o LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.2 SAFETY VALVES Shutdown................................................. 3/4 4-7 Operating............. 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 4-12 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTE0 DURING INSERVICE INSPECTION............................. 3/4 4-17 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION....................... 3/4 4-18 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems,............................... 3/4 4-19 Operational Leakage....................................... 3/4 4-20 -TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...... '3/4 4-22 3/4.4.7 CHEMISTRY,............................................... .3/4 4-24 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS............... 3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SUR REQUIREMENTS...............,..................VEILLANCE 3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY........................................ 3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITV LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY->.1 pCi/ gram DOSE EQUIVALENT I-131..... 3/4 4-29 TABLE 4.4-4 REACTOR: COOLANT SPECIFIC ACTIVITY SAMPLE AN PR0 GRAM.......................................D ANALYSIS 3/4 4-30 3/4,.4.9 PRESSURE / TEMPERATURE LIMITS Reactor' Coolant. System................................... 3/4 4-32 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 16 EFPY................................. 3/4 4-33 FIGURE 3.4-3 REACTOR COOLANT. SYSTEM COOLDOWN LIMITATIONS - -APPLICABLE UP TO 16 EFPY.............................. 3/4 4. TABLE 4.4-5: REACTOR. VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE...................................... 3/4 4 Pressurizer.............................................. 3/4 4-36' Overpressure Protection Systems.......................... 3/4 4-37 3/4.4.10 STRUCTURAL-INTEGRITY..................................... 3/4 4-39 3/4.4.11 REACTOR COOLANT SYSTEM VENTS............................. 3/4 4-40 + CATAWBA - UNITS 1 & 2 VII

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE -- 3/4. 5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Cold Leg-Injection....................................... 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T > 350*F........................... 3/4 5-5 1 avg 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350F........................... 3/.4'5-9 3/4.5.4 . REFUELING WATER STORAGE TANK............................. 3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY. CONTAINMENT Containment Integrity................m 3/4 6-1 Containment Leakage...................................... 3/4 6-2 TABLE 3.6-1 SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS........... 3/4 6-5 Containment Air Locks.................................... 3/4 6-8 Internal Pressure........................................ 3/4 6-10 Air Temperature.............................. 3/4 6-11 . Containment Vessel Structural Integrity..................- 3/4 6-12 Reactor Building _ Structural Integrity.................... 3/4 6-13 Annulus Ventilation System............................... 3/4 6-14 Containment Purge Systems................................ 3/4 6-16 3/4.6.2' OEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System................................. 3/4 6-18 3/4.6.3 CONTAINMENT ISOLATION VALVES............. 3/4 6-20 TABLE ~3.6-2 CONTAINMENT ISOLATION VALVES.......................... 3/4 6-22 3/4.6.4-COMBUSTIBLE GAS CONTROL -Hydrogen Monitors........................................ 3/4 6-38 Electric Hydrogen Recombiners............................ 3/4 6-39 Hydrogen Mitigation System............................... 3/4 6 3/4.6.5 ICE CONDENSER Ice Bed.................................................. 3/4 6-41 Ice Bed Temperature Monitoring System.................... 3/4 6-43 Ice Condenser Doors...................................... 3/4 6-44 Inlet-Door Position Monitoring System.................... 3/4 6-46 4 CATAWBA - UNITS 1 & 2 VIII Amendment No. 86 Amendment No. 80 ((Unit 1) Unit 2) ~

BASES SECTION PAGE 3/4.0 -APPLICABILITY............................................... B 3/4 0-1 3/4.1-REACTIVITY CONTROL SYSTEMS-3/4.1.1 BORATION CONTR0L.......................................... B 3/4 1-1 -3/4.1.2 BORATION SYSTEMS................. Bi/41-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................ B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS (Unit 1).......................... B 3/4 2-1 3/4.2.1 - AXIAL FLUX DIFFERENCE (Unit 1)............................ B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY-RISE HOT CHANNEL FACTOR (Unit 1)................. B 3/4.2-1 3/4.2.4 -QUADRANT POWER TILT RATIO (Unit1)........................ B 3/4 2-3 3/4.2.5-DNB PARAMETERS (Unit 1)............... .B-3/4 2-3 3/4.2 POWER DISTRIBUTION LIMITS (UNIT 2)........................ B 3/4 2-5

3/4.2.1-AXIAL FLUX DIFFERENCE (Unit 2)............................

B 3/4 2-5 '3/4.2.2 and.3/4.2.3 HEAT FLUX HOT ~ CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT-CHANNF' FACTOR (Unit 2).......................... B 3/4 2-7 13/4.2.4 QUADRANT POWER TILT RATIOL(Unit2)........................ B 3/4 2-9 L 3/4.2.5 DNB PARAMETERS (Unit 2)................................... B 3/4 2-10 3/4.3 : INSTRUMENTATION 3/4.3.1' and-3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM-INSTRUMENTATION...............B 3/4 3-1 23/4.3.3 : MONITORING INSTRUMENTATION................ B 3/4 3-3 3/4.3.4' TURBINE'0VERSPEED PROTECTION.............................. B 3/4 3-7 = 1. CATAWBA UNITS 1 & 2 -XIIIe Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

BASES SECTION PAGE 3/4.4 -REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION............. B 3/4 4-1 3/4.4.2 SAFETY VALVES.......................................... B 3/4 4-1 3/4.4.3 PRESSURIZER............................... B'3/4 4-2 3/4.4.4 RELIEF VALVES............................................. B 3/4 4-2 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-2 3/4.4.6 REACTOR C00LAhT SYSTEM LEAKAGE........................... B 3/4 4-3 1 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 l l CATAWBA - UNITS 1 & 2 XIIIb Amendment No. 86 (Unit 1) l l Amendment No. 80 (Unit 2) ) 1 n

v 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 -The-combination of THERMAL POWER, pressurizer pressure, and the highest operating. loop coolant temperature (T,yg) shall not exceed the limits shown in Figu're 2.1-la'(Unit 1) and 2.1-lb (Unit 2) for four loop operation. [ APPLICABILITY: MODES 1 and 2. ACTION: Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT-STANDBY within 1 hour, and comply with the require-ments of Specification 6.7.1. REACTOR COOLANT SYSTEM PRESSURE 2.1.2-The Reactor Coolant System pressure shall not exceed 2735 psig. APPLICABILITY: MODES 1, 2, 3, 4, and 5. ACTION: MODESL1 and 2: Whenever the Reactor Coolant System pressure has exceeded 2735 psig,-be-in HOT-STANDBY _with-the-Reactor Coolant System pressure within its limit within 1 hour, and comply with the requirements of Specification 6.7.1. MODES 3,-.4, and 5: Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor. Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1. CATAWBA - UNITS 1 & 2 2-1 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2) t P 4 v +t 'aw=-" w *d 4 It- !1 e-1 r F N e "r~" e*-7 r r-E+ e

m. 665 660 Unacceptable 655* Operation 650 2400 psla 645 2250 pais 640-i 635 _ 630' I E p625 2000 pois t i m co 620; o C 615; 610 - 1915 psia 605 600i Acceptable 595j Operation 590 + 585t I 580 0.0 0.2 0.4 0.6 0.8 10 1.2 Fraction of Rated Thermal Power FIGURE 2.1-la REACTOR CORE SAFETY LIMITS - FOUR LOOPS IN OPERATION, UNIT 1 s CATAWBA - UNITS 1 & 2 2-2 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

t 660 \\ ; j I N I I I i g j UNACCEPTABLE -~ j OPER ATION As! l l N I iI Ai j N4,00 PSIA l i I 'I h l I X ! l-I I I I I 640 I i i A i I l% i f i I I I I %2250 PSIA l N l l-V i i i i N i l I \\! N i i i N I i \\ l N i i i i i N .h -\\ i. I I Ni I i i i i-N \\ - l.- 1 i i A l i j \\ o' I I I I h I I I '\\ \\

620 i

i i %2000 PSIA } \\ \\ i I i N i \\i\\ [ ! N t i I i N i .i-N' I N i I i'NL i t \\ a: I Ni t i N -\\ l-I h I "N m i i I A i i l-j j l j N1775 PSIA. \\i I I i l N x\\- ~- i I I i N l\\ I-i i i N. il I i l I N I AdCEPTABLE OPERATION. h ~ l I h i I l 580 j I I I l I o 0.2 0.4 0.6 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER FIGURE 2.1-lb REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION, UNIT 2 s CATAWBA - UNITS 1 & 2 2-2a Amendment No. 86 (Unit 1) l Amendment No. 80 (Unit 2)

k.

TABLE 2.2.-1 h; REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS c TOTAL SENSOR 1 z ALLOWANCE -ERROR 3 FUNCTIONAL UNIT (TA) Z_ (S) TRIP SETPOINT ALLOWA8LE VALUE 's. 1. Manual Reactor Trip. N.A. N.A. N.A. N.A. N.A. [ 2. Power Range, Neutron Flux a. High-Setpoint 7.5 5.92 0 $109% of RTP* <110.9% of RTP* b. Low Setpaint 8.3 5.92 0 $25% of RTP* 127.1% of RTP* 3. Power Range, Neutron' Flux,

1. 6 0.5 0

<5% of RTP* with <6.3% of RTP* with High Positive Rate' a time constant a time constant 3 2 seconds 1 2 seconds 4. Power Range, Neutron Flux,

1. 6 0.5 0

$5% of'RTP* with $6.3% of RTP* with High Negative Rate a time constant a time constant 22 seconds 12 seconds 1 5. Intermediate Range, 17.0 8.4 0 $25% of RTP* $31% of RTP* Neutron Flux 6. Source Range, Neutron Flux. 17.0 10 0 $105 cps <1.4 x 105 cps 7. Overtemperature AT 6.98 3.0 2.12 See Note 1 See Note 2 8. Overpower AT 4.9 1.24

1. 7

.See Note 3 See Note 4 (( 9. Pressurizer Pressure-Low. 4.0 2.21 1.5 21945 psig 11938 psig*** 10. Pressurizer Pressure-High-

7. 5 0.71 0.5

$2385 psig $2399 psig i i 11. Pressurizer Water Level-High 5.0 2.18

1. 5

$92% of instrument '<93.8% of instrument span -span PP 12. Reactor Coolant Flow-Low 2.92 1.48 0.6 >90% of loop >88.9% of loop i a> a>. minimum measured iinimum measured flow"* flow ** 22 oo 37 - "RTP = RATED THERMAL POWER

- Loop minimum measured flow = 96,900 gpa (Unit 2), 96,250 gpa (Unit 1) l mw
  - Time constants utilized in the. lead-lag controller for Pressurizer Pressure-Low are 2 seconds for lead.

and.1 second for lag. Channel calibration shall ensure that these time constants are adjusted to these values. l t

i 'y n Q. TABLE 2.2-1 (Continued) E TABLE NOTATIONS >~ NOTE 1: OVERTEMPERATURE AT AT (1-+'t,5) (1 + TsS} I, (1 + T 1 5) 1 2 2 (1,1 5) U (1 + 1 g

5) - T'] + K (P - P') - f (AI)}

(7.+ 1 S) o 1 - 2 3 3 8 "e -m Where: AT Measured AT by Loop Narrow Range RTDs; = g 1+t y ,'Sg lead-lag compensator on measured AT; = m Time constants utilized in lead-lag compensator for AT, ti = 12 s, T1. T2 = T2 = 3 s; l 1 6 1+T 5 Lag c a ensator on measured AT; ~ = 3 Time constant utilized in the lag compensator for AT,1 r3 = 3 = 0; n4 AT, Indicated AT at RATED THERMAL POWER; = K = i 1.38; 't j. 4 K = 2 0.02401/*F; I 1 * ** 7 The function generated by the lead-lag compensator for T = g dynamic compensation; avg h Time constants utilized in the lead-lag compensator for T,yg, 1 14,.Is = 4 = 22 s, .l ,g is = 4 s; r* <+ y T Average temperature, *F; = ' E$ 1 + tsS Lag c spensator on measured T,yg; ] = nncc hh ' Time constant utilized in the measured T,yg lag compensator, is = 0; I Te = mw vu ~

n3 TABLE 2.2-1 (Continued) >6 TABLE NOTATIONS (Continued) + NOTE 1: (Continued) C = 3 T' 590.8*F'(Nominal T,yg allowed by Safety Analysis); Ka 0.001189; = g e-P = Pressurizer. pressure, psig; y P' 2235 psig (Nominal RCS operating pressure); = S = Laplace transform operahr, s 1; and f (AI) is a function of the indicated difference between top and bottom detectors of the 3 power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant STARTUP tests such that: y (i) For q 9 between -22. 5% and -6.5%, t b f (AI) = 0, where.qt and q are percent RATED THERMAL POWER in the top and bottom b halves of the core respectively, and gt*9b is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) For each percent that the magnitude of q a is more negative than -22.5%, the ~ t b (( AT Trip Setpoint shall be automatically reduced by 3.151% of its value at RATED l k"E ' THERMAL POWER; and. 2 55 (iii) For each percent that the magnitude of qt ~9 is a re positive than -6.5%, the AT Trip b yy Setpoint shall be automatically reduced by 1.641% of its value at RATED THERMAL POWER.. NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed. Trip setpoint by mm (( more than 3.0%. l-en fo N vv l

i-m n 2" TABLE 2.2-1 (Continued) @E TABLE NOTATIONS (Continued) NOTE 3: (Continued) f 0.001707/*F for T > 590.8"F and Ks = 0 for T $ 590.8'F, Ks = v. As defined in Note 1, T = e Indicated T,yg at RATED THERMAL POWER (Calibration temperature for AT T" = n, instiumentation, 5 590.8'F), As defined in Note 1, and S = 0 for all al. f (AI) = 2 NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by j q> more than 2.8%. l o l 1 l r* r* l l c> o> C3 Ch l C =? r r+ $I I i l .=

2.1 SAFETY t!MITS BASES 2.1.1 REACTOR CORE (FOR UNIT 1) 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 (ONB) 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 BWCMV correlation. The BWCMV DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio, (DNBR), is defined as the retio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB. The DNB design bssis is as follows: there must be at least a 95% probability that the minimum DNBR of the limiting rod during Condition I and 11 events is greater than or equal to the DNBR limit of the DNB correlation being used (the BWCMV correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data 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, fuel fabrication parameters, and in the BWCMV DNB correlation 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 parameters are used to deter-mine the plant DNBR uncertainty. This DNBR uncertainty is used to establish 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 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 is less than the enthalpy of saturated liquid. l CATAWBA - UNITS 1 & 2 B 2-1 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

2,1 SAFETY LIMITS BASES These curves are based on a nuclear enthalpy rise hot channel f actor. N FaH, of 1.49 for Westinghouse Optimized Fuel Assemblies (0FA's) and 1.55 for the BWFC Mark-BW Fuel Assemblies and a reference cosine with a peak o( 1.55 for axial power shape. An allowance is included for an increase in F at reduced power based on the expression: 3g N F3g = 1.49 (1 + 0.3 (1-P)] for the Westinghouse 0FA's l N FaH = 1.55 (1 + 0.3 (1-P)] For the BWFC Mark-BW's 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 f (al) function of the Overtemperature AI trip. When the axial power imbalance { is not within the tolerance, the axial power imbalance ef feet on the Over-temperature AT trips will reduce the Setpoints to provide protection consistent with core Safety Limits. 2.1.1 REACTOR CORE (FOR UNIT 2) l The restrictions of this Safety Limit ?revent 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 bect.use of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in h'at transfer e 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 predict the ONB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio, (DNBR), is defined as the ratio of the heat flux that would cause DNB at a particular cork location to the local heat flux, and is indicative of the margin to DNB. The ONB 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 11 events is greater than or equal to the DNBR limit of the ONB correlation being used (the WRB-1 correlation in this application). The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95% probability with 95% confidence that DNB will not occur when the minimum DNBR is at the ONBR limit. CATAWBA - UNITS 1 & 2 B 2-2 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

2.1 SAFETY LIMITS BASES 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% c.onfidence that the minimum DNBR for the limiting rod is greater than or eaval to the ONBR limit. The uncertainties in the above plant parameters are used to determine the plant ONBR 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 ir.put parameters without uncertainties. The curves of Figure 2.1-1 show the loci of points of THERMAL p0WER, Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the der,ign DNBR value, or the average enthalpy at the vessel exit is less than the e'ithalpy of saturated liquid. This curve is beled on a nucleer enthalpy rise hot channel factor, N F3g, of 1.49 and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in F at reduced power based on the expression: g N F r 3g 1.49 [1 + 0.3 (1-P)) Where p is the fraction of RATr.D THERHAL 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 imaalance is within the limits of the f (al) function of the Overtemperature t'ip. When the axial power imbalance is not within the tolerance, the axial power imbalance ef fect on the Over-temperature aT trips will rtduce 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 radionuclides contained in the reactor coolant from reaching the containment atmosphere. The reactor vessel, pressurizer, and the Reactor Coolant System piping, valves, and fittings 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 125% (3110 psig) of design pressure, to demonstrate integrity prior to initial operation. CATAWBA - UNITS 1 & 2 8 2-2a Amendment No. 86 (Unit 1) l Amendment No. 80 (Unit 2)

UN!? 1 l 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FlVX DIFFERENCE (AFO) LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the acceptable limits specified in the CORE OPERATING LIMITS REPORT (COLR). APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER.* (Unit 1) ACTION: For operation with the indicated AFD outside of the limits specified l a. in the COLR, 1. Either restore the indicated AFD to within the COLR 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 limits specified in the COLR.

See Special Test Exceptions Specification 3.10.2.

I CATAWBA - UNITS 1 & 2 3/4 A2-1 Amendment No. 86 (Unit 1) l Amendment No. 80 (Unit 2)

- _. _ = _ - UNIV 1 l POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION bove 50% of RATED THERMAL POWER by: Monitoring the indicated AFD for each OPERABLE excore channel: a. 1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2) At leant once per hour for the first 24 hours after restoring the AFD Monitor Alarm to OPERABLE status. b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoper-able. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging. c. The provisions of Specification 4.0.4 are not applicable. 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. CATAWBA - UNITS 1 & 2 3/4 A2-2 l AmendmentNo.b(Unit 1) Amendment No, (Unit 2)

UNIT 1 1 POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F (X,Y,Z) l LIMITING CONDITION FOR OPERATION 3.2.2 F (X,Y,Z) shall be limited by imposing the following relationships: 9 I f "^(X,Y,Z) $ F # q Q K(Z) for P > 0.5 P .,4 MA(X,Y,Z) $ RTP F F g Where: FRTP = the F Limit at RATED THERMAL POWER q q specified in the CORE OPERATING LIMITS REPORT (COLR), MA(X,Y,Z) = the measured heat flux hot channel factor F F N O with adjustments as specified in 4.2.2.3, q (X,Y.Z), p., THERMAL POWER , and - RATED THERMAL POWER K(Z) = the normalized F (X,Y,Z) limit specified in the 9 COLR for the appropriate fuel types. APPLICABILITY: MODE 1. (Unit 1) ACTION: With F (X,Y,Z) exceeding its limit: 9 Reduce THERMAL POWER at least 1% for each 1% Fg (X,Y,Z) exceeds the MA a. limit within 15 minutes and similarly' reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours, and b. Control the AFD to within new AFD limits which are determined by reducing the allowable power at each point along the AFD limit lines of Specification 3.2.1 at-least 1% for each 1% Fq (X,Y,Z) exceeds the limit within 15 minutes and reset the AFD alarm setpoints to the modified limits within 8 hours, and POWER OPERATION may proceed for up to a total of 72 hours; subsequent c POWER OPERATION may proceed provided the Overpower AT Trip setpoints (value of K ) have been reduced at least 1% (in ai span) for each 1% 4 M(X,Y,Z) exceeds the limit, and fq d. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may then be increased provided F (X,Y,Z) is I 9 demonstrated througrfincore niapping to be within its limit. CATAWBA - UNITS 1 & 2-3/4 A2-3 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

UNIT 1 } POWER O!STRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. M(X,Y,ZhldhallbeevaluatedtodeterminewhetherF(X,Y,Z) 4.2.2.2 F iswithinktslimitby; 9 Using the movable incore detectors to obtain a power distributNn a. map at any THERMAL POWER greater than 5% of RATED THERMAL POWER. M b. Measuring Fq (X,Y,Z) at the earliest of: 1. At least once per 31 Effective Full Power Days, or 2. Upon reaching equilibrium conditions after exceeding by 10% or more of RATED THERMAL POWER, the THERMAL POWER at which M Fq (X,Y,Z) was last determined (2), or 3. At each time the QUADRANT POWER TILT RATIO indicated by the excore detectors is normalized using incore detector measurements. c. Performing the following calculations: 1. For each location, calculate the % margin to the maximum allowable design as follows: M F(X,Y,2) % Operational Margin = (1 - g ) x 100% [Fh(X,Y,Z)]oP M F (X,Y,Z) %RPSMargin=(1-n ) x 100% b [Fh(X,Y,7 L OP L RPS where[F(X,Y,Z)] and[F(X,Y,2)) are the Operational and l 9 n RPS design peaking limits defined in the COLR. 2. Find the minimum Operational Margin of all locations examined in 4.2.2.2.c.1 above. If any margin is less than zero, then either of the following actions shall be taken: WNo additional uncertainties are required in the following equations for M-Fq (X,Y,Z), because the limits include uncertainties. E)Duringpowerescalationatthebeginningofeachcycle,THERMALPOWERmay be increased until a pow n level for extended operation has been achieved and a power distribution map obtained. CATAWBA - UNITS 1 & 2 3/4 A2-4 Amendment No. 86 Unit 1 ( Amendment No. 80 Unit 2

UNIT 1 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) (a) Within 15 minutes: (1) Control the AFD to within new AFD limits that are determined by: C) reduced COLR (AFD Limit) negative = (AFD Limit) negative ~ min + [NSLOPE:(3) x Margin 0P] absolute value C) reduced COLR (AFD Limit)po,gggy, = (AFD Limit)po,4ggy, min -[PSLOPE (a) x Margin 0P] absolute value min where Margin is the minimum margin from 4.2.2.2.c.1,$$d (2) Within 8 hours, reset the AFD alarm setpoints to the modified limits of 4.2.2.2.c.2.a or (b) Comply with the ACTION requirements of Specification 3.2.2. 3. Find the minimum RPS Margin of all locations examined in 4.2.2.2.c.1 above. If any margin is less than zero, then the following action shall be taken: Within 72 hours, reduce the K t value for OTAT by: adjusted = K ( ) - [KSLOPEI ) x Margin 3] absolute value Kt min where MARGIN is the minimum margin from 4.2.2.2.c.1. RPS 0 ) Defined and specified in the COLR per Specification 6.9.1.9. 0)K value from Table 2.2-1. 1 l CATAWBA - UNITS 1 & 2 3/4 A2-5 Amendment No. 8 AmendmentNo.8b(Unit 1) (Unit 2)

UNIT 1 POWER DISTRIBUTION LIMITS j SURVEILLANCE REQUIREMENTS (Continued) Extrapolating the two most recent measurements to 31 Effective Full d. Power Days beyond the most recent measurement and if: M L OP -[F (X,Y,Z) (extrapolated) 1 [F (X,Y,Z) (extrapolated),or 9 g M L RPS [F (X,Y,Z) (extrapolated) 3 [F (X,Y,Z] (extrapolated), g 9 either of the following actions shall be taken: M 1. F'(X,Y,Z) shall be increased by 2 percent over that specified in i 0 4.2.2.2.a. and the calculations of 4.2.2.2.c repeated, or 2. A movable incore detector power distribution map shall be obtained, and the calculations of 4.2.2.2.c.1 shall be performed no later than the time at which the margin in 4.2.2.2.c.1 is extrapolated to be equal to zero, The limits' in Specifications 4.2.2.2.c and 4.2.2.2.d are not applicable e. in the following core plane 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. 4.2.2.3-When a full core power distribution map is taken for reasons other than meeting the requirements of Specification 4.2.2.2, an overall_Fq (X,Y,Z) M shall be determined, then increased by 3% to account for manufacturing tolerances, further increased by 5% to account for measurement uncertainty, and further increased by the radial-local peaking factor to obtain a maximum local peak. TMs value shall be compared to the limit in Specification 3.2.2. r CATAWBA - UNITS 1 & 2 3/4 A2-6 Amendment No.86 (Unit 1) Amendment No. 80 (Unit 2)

UNIT 1 POWER DISTRIBUTION LIMITS i 3/4.2.3 NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR FAH(X,Y) LIMITING CONDITION FOR OPERATION 3.2.3 F3g(X,Y)shall be limited by imposing the following relationship: M l FAHR (X,Y) < FAHR (X,Y) N Where: FAHR (X,Y) = the maximum measured radial peak ratio as defined in the CORE OPLRATING LIMITS REPORT (COLR). l FAHR (X,Y) = the maximum allowable radial peak ratio as defined in the (COLR). APPLICABILITY: MODE 1. (UNIT 1) ACTION: With F3g(X,Y) exceeding its limit: Within 2 hours, reduce the allowable THERMAL POWER from RATED THERMAL a. I ) for each 1% that FAHR (X,Y) exceeds the limit, and N POWER at least RRH% b. Within 6 hours either: M 1. Restore FAHR (X,Y) to within the limit of Specification 3.2.3 for RATED THERMAL POWER, or 2. Reduce the Power Range Neutron Flux-High Trip Setpoint in Table 2.2-1 at least RRH% for each 1% that FAHR"(X,Y) exceeds that limit, and Within 72 hours of initially being outside the limit of Specification c. 3.2.3, either: N 1. Restore FAHR (X,Y) to within the limit of Specification 3.2.3 for RATED THERMAL POWER, or 2. Perform the following actions: (a) Reduce the OTAT K term in Table 2.2-1 by at least TRHI) i N for each 1% that FAHR (X,Y) exceeds the limit, and N (b) Verify through incore mapping that FAHR (X.Y) is restored to within the limit for the reduced THERMAL F 't allowed by ACTION a, or reduce THERMAL PC'wTr i 5% of RATED THERMAL POWER within the next 2 riours. C)RRH is the amount of THERMAL POWER rgduction required to compensate fo M each 1% that FAHR (X,Y) exceeds FAHR (X,Y) provided in the COLR per Specification 6.9.1.9. ( )TRH is the amount of OTAT K g i setpoint reduction required to compensate for each 1% that FAHR (X,Y) exceeds the limit of Specification 3.2.3, provided in the COLR per Specificatidn 6.9.1.9. . CATAWBA - UNITS 1 & 2 3/4 A2-7 Amendment No. 86 (Unit 1) An;?ndment No. 80 (Unit 2)

. _.. _ _ _ _ _ _ _ _ _ _ _ _. _. _. _ _ _ ~ UNIV 1 } POWER DISTRIBUTION tIMITS i LIMITING CONDITION FOR OPERATION = ACTION (Continued) d. 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. and/or c.2., above; subsequent POWER OPERA-N TION may ptoceed provided that FAHR (X,Y) is demonstrated, through incore flux mapping, to be within the limit specified in the COLR prior to exceeding the following THERMAL POWER levels: 1) 50% of RATED THERMAL POWER, 2) 75% of RATED THERMAL 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. N 4.2.3.2 FAHR (X,Y) shall be evaluated to determine whether F3g(X,Y) is within its limit by: Using the movable incore detectors to obtain a power distribution a. map at any THERMAL POWER greater than 5% of RATED THERMAL POWER. M b. Measuring FAHR (X,Y) according to the following schedule: 1. Prior to operation above 75% of RATED THERMAL POWER at the beginning of each fuel cycle, and the earlier of: 2. At least once per 31 Effective Full Power Days, or 3. At each time the QUADRANT POWER TILT RATIO indicated by the excore detectors is normalized using incore detector measurements. c. Performing the following calculations: 1. For each location, calculate the % margin to the maximum allowable design as follows: M -%F Margin = 1 - FAHR (X,Y)) x 100% g FAHR'(X,Y) M Noadditiona[(X,Y),includesuncertaintiesuncertainities are required for F because FAHR CATAWBA - UNITS 1 & 2 3/4 A2-8 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

UNIT 1 l POWER OISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued) 2. Find the minimum margin of all locations examined in 4.2.3.2.c.1 above. If any margin is less than zero, comply with the ACTION requirements of Specification 3.2.3. d. Extrapolating the two most recent measurements to 31 Ef fective full Power Days beyond the most recent measurement and if: FaHRM (extrapolated) > FaHRL (extrapolated) either of the following actions shall be taken: M 1. FAHR (X,Y) shall be increased by 2 percent over that specified in 4.2.3.2.a, and the calculations of 4.2.3.2.c repeated, or 2. A movable incore detector power distribution map shall be obtained, and the calculations of 4.2.3.2.c shall be performed no later than the time at which the margin in 4.2.3.2.c is extrapolated to be equal to zero. l .s CATAWBA - UNITS 1 & 2 3/4 A2-9 Amendment No. 86 (Unit 1) l Amendment No. 80 (Unit 2)

UNIT-1 l POWER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER (Unit 1).*,** ACTION: With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but a. 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 redcced to within its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. 2. Within 2 hours either: a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.02 and similarly reduce the Power Range Neutron l Flux-High Trip Setpoints within the next 4 hours. 3 Verify that the' QUADRANT POWER TILT RATIO is within its limit within 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 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 once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.

See Special Test Exceptions Specification 3.10.2.
- Not applicable until calibration of the excore detectors is completed subse-quent to refueling, CATAWBA - UNITS 1 & 2 3/4 A2-10 Amendment No. b ((Unit 1) l Amendment No, Unit 2)

... - _ ~. _... _ _. _, _. -. _ - _ _... = --.~ - -

UNIT 1 l POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION A_CTION (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 tour 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. 2. Reduce THfRMAL POWER at least 3% from RATED THERMAL POWER for each 1% 01 indicated QUADRANT POWER TILT RATIO in excess of i 1.02, within 30 minutes; I 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 5etpoints 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 once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER, 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. CATAWBA - UNITS 1 & 2 3/4 A2-11 Amendment No. 86 (Unit 1) l Amendment No. 80 (Unit 2)

UNIT 1 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 POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified at 95% or greater RATED THERMAL POWER. d. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE kEQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by: Calculating the ratio at least once per 7 days when the alarm is a. OPERABLE, and b. Calculating the ratio at least once per 12 hours during steady-state operation when the alarm is inoperable, The provisions of Specification 4.0.4 are not applicable. c. 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 full-cora flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours. s CATAWBA - UNITS 1 & 2 3/4 A2-12 Amendment No. 86 (Unit 1) ( Amendment No. 80 (Unit 2)

UN!? 1 l . POWER DISTRIBUTION LIMITS l 3/4.2.5 DNB PARAMETERS 1 LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1: Reactor Coolant System T,yg, a. b. Pressurizer Pressure, c. Reactor Coolant System Total Flow Rate. APPLICABILITY: MODE 1. (Unit 1) ACTION: With either of the parameters identified in 3.2.5a. and b. above a. 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. b. With the combination of Reactor Coolant System total flow rate and THERMAL POWER within the region of restricted operation specified on Figure 3.2-1, within 6 hours reduce the Power Range Neutron Flux-High Trip Setpoint to below the nominal setpoint by the same amount (% RTP) as the power reduction required by Figure 3.2-1. With the combination of Reactor Coolant System total flow rate and c. THERMAL POWER within the region of prohibited operation specified on Figure 3.2-1: 1. Within 2 hours either: a) Restore the combination of Reactor Coolant System total flow rate and THERMAL POWER to within the region of permissible operation, or b) Restore the combination of Reactor Coolant System total flow rate and THERMAL POWER to within the region of restricted operation and comply with action a. abovc, or c) Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours. i CATAWBA - UNITS 1 & 2 3/4 A2-13 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

UNIT 1 l , POWER OISTRIBUTION LIMITS 3/4.2.5 DN8 PARAMETERS LIMITING CONDITION FOR OPERATION 2. Within 24 hours of initially being within the region of prohibited operation specified on Figure 3.2-1, verify that the combination of THERMAL POWER and Reactor Coolant System total flow rate are restored to within the regions of restricted or permissible operation, or reduce THERHAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours. SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours. 4.2.5.2 The Reactor Coolant System total flow rate indicators shall be sub-jected to a CHANNEL CALIBRATION at least once per 18 months. The measurement instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow measurement. 4.2.5.3 The Reactor Coolant System total flow rate shall be determined by precision heat balance measurement at least once per 18 months. i t a l CATAWBA - UNITS 1 & 2 3/4 A2-14 Amendment No. 86 (Unit 1) l Amendment No. 80 (Unit 2) i l \\

UNIT 1 \\ TABLE 3.2-1 DNB PARAMETERS PARAMETER LtHITS Four Loops in Operation Average Temperature Meter Average - 4 channels: < 592'F - 3 channels: {592'F Computer Average - 4 channels: < 593'F - 3 channels: [593'F Pressurizer Pressure Meter Average - 4 channels: > 2227 psig* - 3 channels: > 2230 psig* Computer Average - 4 channels: > 2222 psig* - 3 channels: [2224psig* Reactor Coolant System Total Flow Rate figure 3.2-1

Limit not applicable during either a THERMAL POWER ramp in excess of 5% of l

RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER. CATAVBA - UNITS 1 & 2 3/4 A2-15 Amendment No. 86 (Unit 1) l Amendment No. 80 (Unit 2)

UNIT 1 l 3sseso A penalty of o.1% for unestectos f.eewater ventwt fW and a Permissible j i j mesewoment scortanty of Li% for Operation no. er.== n m. on.. g,,,,, I 1 a s 5 00 0...................................................... 9,sj,s sooo). I 1 I I l Restricted Operation (ss.3suso) E 3811501 ,5 q, Region 2 l E f 4 3 I g 3773ng. (sd.37730o) LA. i v3 I'E3 * *'I I e 373450< O g Prohibited g 4 Operation g369600 1 t 388750 381900 84 88 90 92 94 94 94 10 0 102 i Fraction of Rated Thermal Power i ( Figure 3.2-1 Reactor Coolant System Total Flow Rate Versus Rated Thermal Power - Four Loops in Operation (Unit 1) CATAWBA - UNITS 1 & 2 3/4 A2-16 Amendment No. 86 (Unit 1) I Amendment No. 80 (Unit 2)

_ ~_.. - _ - UNIT 2 l 3f4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AF0] LIMITING CCN0! TION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within: the allowed operational space as specified in the CORE OPERATING a, LIMITS REPORT (COLR) for RA0C operation, or b. within the target band specified in the COLR about the target flux difference during baseload operation. APPLICABILITY: MODE 1, above 50% of RATED THERMAL POWER * (Unit 2) ACTION: For RAOC operation with the indicated AFD outside of the limits a. specified in the COLR, 1. Either restore the indicated AFD to within the COLR limits within 15 minutes, or 2. Reduce THERW L 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, ND** b. For Base Load operation above APL with the indicated AXIAL FLUX DIFFERENCE outside of the applicable target band about the target flux difference: 1. Either restore the indicated AFD to within the COLR specified target band limits within 15 minutes, or ND 2. Reduce THERMAL POWER to less than APL of RATED THERMAL POWER and discontinue Base Load operation within 30 minutes, c. THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR. j

See Special Test Exceptions Specification 3.10.2.

ND

- APL is the minimum allowable (nuclear design) power level for base load operation and is specified ip the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9.

CATAWBA - UNITS 1 & 2 3/4 B2-1 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2) l l

UNIT 2 l _ POWER DISTRIBUTION LIMITS _ LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENTS 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: Monitoring the indicated AFD for each OPERABLE excore channel: a. 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 af ter restoring the AFD Monitor Alarm to OPERABLE status. b. Monitoring and logging the indicated AFD for each OPERABLE excore channel at least once per hour for the first 24 hours and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoper-able. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging, The provisions of Specification 4.0.4 are not applicable. c. 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. 4.2.1.3 When in Base Load operation, the target axial 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. 4.2.1.4 When in Base Load operation, the target flux difference shall be updated at least once per 31 Effective Full _ Power Days by either determining the target flux difference in conjunction with the surveillance requirements of Specification 3/4.2.2 or by linear interpolation between the most recently mea-sured values and the calculated value at the end of cycle life. The provisions of Specification 4.0.4 are not applicable. l l e CATAWBA - UNITS 1 & 2 3/4 B2-2 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

UNIT 2 l POWER DISTRIBUTION LIMITS 3/4;2.2 HLATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships: g J RTP F (Z) 1 F g P RTP i F (Z) i 9 K(Z) for P s 0.5 Where: F TP = the F Limit at RATED THERMAL POWER (RTP) g specified in the CORE OPERATING LIMITS REPORT (COLR), p, THERMAL POWER , and RATED THERMAL POWER K(Z) = the normalized F (Z) for a given core height 9 specified in the COLR, APPLICABILITY: MODE 1. (Unit 2) ACTION: With F (Z) exceeding its limit: 9 a. Reduce THERMAL POWER at least 1% for each 1% F (Z) ex;ewds the limit g within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours; POWER OPERATION may proceed for up to a total of 72 hours; subsequent POWER OPERATION may proceed provided the Overpower Al Trip Setpoints (value of K ) have been reduced at least 1% (in AT span) for each 1% F (Z) exceeds the limit, and q b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a., above; THERMAL POWER may-then be increased provided F (Z) is demonstrated through incore mapping to be within its limit. g s CATAWBA - UNITS 1 & 2 3/4 B2-3 Amendment No. 86 (Unit 1). l 3 Amendment No. 80 (Unit 2)

UNIT 2 l POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable. 4.2.S.2 for RAOC operation, F (z) shall be evaluated to determine if F (z) is within its limit by: g 9 Using the movable incore detectors to obtain a power distribution a. map at any THERMAL POWER greater than 5% of RATED THERMAL POWER,. b. Increasing the measured F (z) component of the power distribution n map by 3% to account for manuf6cturing tolerances and further in-creasing the value by 5% to account for measurement uncertainties. Verify the requirements of Specification 3.2.2 are satisfied. Satisfying the following relationship: c. RTP F hz)1 q x g(*) for P > 0.5 F q P x W(z) RTP F hz)1 Q x K(z) for P < 0.5 F q W(z) x 0.5 where F (z) is the measured F (z) increased by the allowances for g manufacturingtolerancesandmeasurementuncertainty,FfTP is the F limit, K(z) is the normalized F (z) as a function of core height, q q P is the relative THERMAL POWER, and W(z) is the cycle dependent function that accounts for power distribution transients encountered TP during normal operation. F , K(z), and W(z) are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9. M J. Measuring F (z) according to the following schedule: q 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 9 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.

CATAWBA - UNITS 1 & 2 3/4 B2-4 Amendment No. 86 (Unit 1) l Amendment No. 80 (Unit 2)

UNIT 2 l POWER DISTRIBUTION LlHITS SURVEILLANCE REQUIREMENTS (Continued) e. With measurements indicating maximum F" (z) 9 over z K(z) has increased since the previous determination of F"(z) either of the following actions shall be taken: 9 M 1) F (z) shall be increased by 2% over that specified in g Specification 4.2.2.2c., or 2) F (2) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that maximum F (z) is not increasing. over z K(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 expression: 9 r -1 ([ maximum F (2) x W(z) h,1 x 100 for P > 0.5 4 over 2 RTP / ~ p ., h

KCZ) )

1 5 f M -) I maximum F (z) x W(z) g ,1 x 100 for P < 0.5 over z RTP 1 x K(z) )! \\ .5

2) 'One 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 Specification 3.2.1 by 1% AFD for each percent F (z) exceeds 9 its limits as determined in Specification 4.2.2.2f.1). Within 8 hours, reset the AFD alarm setpoints to these mod-ified limits, or b) Comply with the requirements of Epecification 3.2.2 for F (z) exceeding its limit by the percent calculated above, or 9 l c) Verify that the requirements of Specification 4.2.2.3 for Base Load cperation are satisfied and enter Base Load operation. CATAWBA - UNITS 1 & 2 3/4 B2-5 Amendment No. 86 (Unit 1) l i Amendment No. 80 (Unit 2)

UNIT 2 l POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (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. 4.2.2.3 Base Load operation is permitted at powers above APLND* if the following conditions are satisfied: PriortoenteringBaseLoadoperation,maintainTHERMALPOWERYbove a. ND APL and less than or equal to that allowed by Specification 4.2.2.2 for at least the previous 24 hours. Maintain Base Load operation surveillance (AFD within the target band about the target flux differ-ence of Specification 3.2.1) during this time period. Base Lnad operation is then permitted providing THERMAL POWER is maintained NU BL ND between APL and APL or between APL and 100% (whichever is most limiting) and FQ surveillance is maintained pursuant to Specification OL 4.2.2.4. APL is defined as: RTP F APL0b = minimum Q x K(Z) [ ] x 100% over Z F (Z) x W(Z)BL where: F (z) is the measured F (z) increased by the allowances for q manufacturing tolerances and measurement uncertainty, FfTPis the F limit, K(z) is the r.ormalized F (2) as a function of core height. q q W(z)gt is the cycle dependent function that accounts for limited power distribution transients encountered during Base Load operation. TP F , K(z), and W(Z)BL are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9. b. During Base Load operation, if the THERMAL POWER is decreased below APL"O then the conditions of 4.2.2.3a shall be satisfied before re-entering Base Load operation. 4.2.2,4 During Base Load Operation F (Z) shall be evaluated to determine if F (Z) is within its limit by: q q Using the movable incore detectors to obtain a power distribution a. ND map at any THERMAL POWER above APL b. Increasing the measured F (Z) component of the power distribution map q by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties. Verify the requirements of Specification 3.2.2 are satisfied. ND

APL is the minimum allowable (nuclear design) power level for Base Load operation in Specification 3.2.1.

CATAWBA - UNITS 1 & 2 3/4 B2-6 Amendment No. 86 (Unit 1) k Amendment No. 80 (Unit 2)

UNIT 3 l _ POWER DISTRIBUTION LIMITS i' SURVEILLANCE REQUIREMENTS (Continued) Satisfying the following relationship: c. RTP N 0 x ND F (Z) $ for P > APL { g p where: F (Z) is the measured F (Z). FhTP t 4, g, p j9,4g, g K(Z) is the normalized F (Z) as a function of core height. P is the 9 -relative THERMAL POWER. W(Z) is the cycle dependent function that accounts for limited power dibribution transients encountered during RTP Base load operation. F , K(Z), and W(Z)BL are specified in the i g CORE OPERATING LIMITS REPORT per Specificatior 6.9.1.9. mination ac$(Z) in conjunction with target flux difference deter-d. Measuring F cording to the following schedule: 1. Frie-to entering Base Load operation after satisfying surveil-lance 4.2.2.3 unlea a full core flux map has been taken in the previous 31 EFPD wi.a the relative thermal power having been ND maintained above APL for the 24 hours prior to mapping, and 2. At least once per 31 effective full power days. e. With measurements indicating M maximum Fg (z) b (z) 3 K over z has increased since the precious determination F (Z) either of the following actions shall be taken: 1. F (Z) shall be increased by 2 percent over that specified in 4.2.2.4c, or

I 2.

F (Z) shall be measured at least once per 7 EFPD until 2 successive. maps indicate that maximum (z) K(z) ) is not increasing. cyer z f. W!th the relationshin specified in 4.2.2.4c above not being satisfied, either of the following actions shall be taken: 1. Placa -the core in an equilibrium condition where the limit in 4.2.2.2c is satisfied, and remeasure F (Z), or c CATAWBA - UNITS 1 & 2 3/4 B2-7 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2) -. ~.

~_ _ _. _ _. _ _ _. _ _ _ _.. _ _ _ _ _ _. _. - _ _ _ _. _ _ _.. _. _. _. UNI? 2 l POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 2, Comply with the requirements of Specification 3.2.2 for F (Z) exceeding its limit by the percent calculated with g the following expression: M F (Z) x W(Z)BL ] ) -1 ) x 100 for P > APL q ND [(max. over z of [ TP F 7 gg7) P g. The limits specified in 4.2.2.4c. 4.2.2.4e., and 4.2.2.4f. above are not applicable in the following core plan regions: 1. Lower core region 0 to 15 percent, inclusive. 2. Upper core region 85 to 100 percent, inclusive. 4.2.2.5 When F (Z) is measured for reasons other than meeting the requirements 9 of Specification 4.2.2.2 an overall measured F (z) shall be obtained from a power g distribution map and increased by 3% to account for manufacturing tolerances and further increas*3d by 5% to account for measurement uncertainty. 1 CATAWBA - UNITS 1 & 2 3/4 B2-8 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

3 UNIT 2 l l POWER DISTRIBUTION LIMITS l 3/4.2.3 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLLAR ENTHA! PY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System total flow rate and R shall be maintained within the region of permissible operation specified in the CORE OPERATING LIMITS REPORT (COLR) for four loop operation. k, Where: N AH a. R= r, F [1.0 + MFaH (1.0 - P)] THERMAL POWER b. P = R TED THERMAL POWER c. F H = Measured values of F H btained by using the movable incore detectors to obtain a power distribution map. The measured values of F shall be used to calculste R sirce tne figure H specified in the COLR includes peralties for undetected feed-water venturi fouling of 0.1% and for measurement uncertainties of 2.1% for flow and 4% for incore measurement of F H' d. FRTP= The F limit at RATED THERMAL POWER (RTP) specified in the H COLR, and e. MF = lhe power factor multiplier specified in the COLR. 3g APPLICABILITY: MODE 1 (UNIT 2). ACTION: a. With the combination of Reactor Coolant System total flow rate ano R within the region of restricted operation within 6 hours reduce the Power Range Neutron Flux-High Trip Setpoint to below the nominal setpoint by the same amount (% RTP) as the power reduction required by the figure specified in the COLR. b. With the combination of Reactor Coolant System total flow rate and R within the region of prohibited operation specified in the COLR: 1. Within 2 hours either: a) Restore the combination of Reactor Coolant System total flow rate and R to within the region of permissible operation, or b) Restore the combination of Reactor Coolant System total flow rate and R to within the region of restricted operation and comply with action a. above, or CATAWBA - UNITS 1 & 2 3/4 B2-9 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

_. _ _.... - _ _ _ _ _._.....______m S V UNIT 2 l POWER DISTRIBUTION LIMITS 3/4.2.3 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION ACTION (Continued) c) Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER withine the next 4 hours. 2.- Within 24 hours of initially being within the region of prohibited operation specified in the COLR, verify through incore flux mapping and Reactor Coolant System total flow rate comparison that the com-bination of R and Reactor Coolant System total flow rate are restored to within the regions of restricted or permissible operation, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours. l s, I CATAWBA - UNITS 1 & 2 3/4 82-10 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

UNIT 2 l POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued) 3. 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 b.1.c) and/or b.2., above; subsequent POWER OPERA-TION may proceed provided that the combination of R cnd indicated Reactor Coolant System total flow rate are demonstrated, through incore flux mapping and Rea: tor Coolant System total flow rate ~ comparison, to be within the regions of restricted or permissible operation specified in the COLR prior to exceeding the following THERMAL POWER levels: a) A nominal 50% of RATED THERMAL POWER, 0) A nominal 75% of RATED THERMAL POWER, and c) Within 24 hours of attaining greater than or equal to 95% of RATED THERMAL POWER. 51RVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable. 4.2.3.2 The combination of indicated Reactor Coolant System total flow rate determined by process computer readings or digital voltmeter measurement and R shall be determined to be within the regions of restricted or permissible operation specified in the COLR: Prior to operation above 75% of RATED THERMAL POWER after each fuel a. loading, and b. At least once per 31 Effective Full Power Days. 4.2.3.3 The indicated Reactor Coolant System total flow rate shall be verified to be within the regions of restricted or permissible operation specified in the COLR at least once per 12 hours when the most recently obtained value of R, obtained per Specification 4.2.3.2, is assumed to exist. 4.2.3.4 The Reactor Coolant System total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months. The measurenient instrumentation shall be calibrated within 7 days prior to the performance of the calorimetric flow measurement. 4.2.3.5 The Reactor Coolant System total flow rate shall be determined by precision heat balance measurement at least once per 18 months. CATAWBA - UNITS 1 & 2 3/4 B2-11 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

~

UNIT 3 l i i POWER DISTRIBUTION LIMITS i 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02. APPLICABILITY: MODE 1, *above 50% of RATED THERMAL POWER (Unit 2) ACTION: With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but a. less than or equal to 1.09: 1. Calculate the QUADRANT POWER LILT 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 THERIAL POWER. 2. Within 2 hours either: a)- Reduce the QUADRANT POWER TILT RATIO to within its liait, or b) Reduce THERMAL POWER at.least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints with % the next 4 hours. 3. Verify that the QUADRANT POWER TILT RATIO is within its' limit within 24 hours after exceeding the limit or reduce THEkMAL 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 RATEC THERMAL POWER may proceed provided that the QUADRANT POWER TIL1 RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95% j. or greater RATED THERMAL POWER. l

See Special Test Exceptions Specification 3.10.2.

CATAWBA - UNITS 1 & 2 3/4 B2-12 Amendment No. 86 (Unit 1) -l Amendment No. 80 (Unit 2) l-

UNIT 2 l 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 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, 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 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 once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER, 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 l b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. l t CATAWBA - UNITS 1 & 2 3/4 B2-13 Amendment No. 86 (Unit 1) l -Amendment No. 80 (Unit 2)

UNIT 2 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 coerect the cause of the out-of-limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified at 95% or greater RATED THERMAL POWER. d. The provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.2.4.1 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by: Calculating the ratio at least once per 7 days when the alarm is a. OPERABLE, and b. Calculating the ratio at least once per 12 hours during steady-state operation when the alarm is inoperable, The provisions of Specification 4.0.4 are not applicable. c. 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 symetric power distribution, obtained from two sets of four symetric thimble locations or full-core flux map, is consistent with the indicated QUADRANT POWER TILT RATIO at least once per 12 hours. CATAWBA - UNITS 1 & 2 3/4 B2-14 Amendment No. g (Unit 1) ) Amendment No. ou (Unit 2) r

- _ _ _. _. _. _.... = _. - _.. UNIT 2- [ POWER OISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS t LIMITING CCNDITION FOR OPERATION 3.2.5 The following DNB 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 Unit 2 only. 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. CATAWBA - UNITS 1 & 2 3/4 82-15 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

~ _ . ~ UNIT 2 TABLE 3.2-1 DNB PARAMETERS PARAMETER LIMITS Four Loops in Operation Average Temperature. Meter Average 4 channels: < 592 F 3 channels: -{592F Computer Average 4 channels: < 593*F 3 channels: 5593*F Pressurizer Pressure Meter Average - 4 channels: 1 2227 psig* - 3 channels: 1 2230 psig* Computer Average - 4 channels: 1 2222 psig* -3 channels: 1 2224 psig* l l -. l l-

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 RATED THERMAL POWER.

4 CATAWBA - UNITS 1 & 2. 3/4 B2-16 Amendment No.86 (Unit 1) Amendment No.80 (Unit 2) . - ~.

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS COLD LEG INJECTION LIMITING CONDITION FOR OPERATION 3.5.1 Each cold leg injection accumulator shall be OPERABLE with: a. The discharge isolation valve open, b. A contained boratec' water volume of between 7704 and 8004 gallons, c. A boron concent 'etween 1900 and 2100 ppm, d. A nitrogen cove, of between 585 and 678 psig, and e. A water level and prersure channel OPERABLE. APPLICABILITY: MODES 1, 2, ano 3^. ACTION: With one cold leg injection accumulator inoperable, except as a result a. of a closed isolation valve or boron concentration less than 1900 ppm, restore the inoperable accumulator to OPERABLE status within 1 hour or be in at least HOT STANDBY 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 isolation valve or be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours. c. With one accumulator inoperable due to boron concentration less than 1900 ppm and: 1) The volume weighted average boron concentration of the three limiting accumulators 1900 ppm or greater, restore the inoperable accumulator to OPERABLE status within 24 hours of the low boron determination or be in at least HOT STANDBY within the next 6 hours and reduce Reactor Coolant System pressure to less than l 1000 psig within the following 6 hours. 2) The volume weighted average boron concentration of the three limiting accumulators less than 1900 ppm but greater than 1500 ppm, restore the inoperable accumulator to OPERABLE status or return the volume weighted average boron concentration of the three limiting accumulators to greater than 1900 ppm and

Reactor Coolant System pressure above 1000 psig.

CATAWBA - UNITS 1 & 2 3/4 5-1 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

9 g EMtRGENCY CORE COOLING SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) enter ACTION c.1 within 6 hours of the low boron determination or be in HOT STANDBY within the next 6 hours and reduce Reactor Coolant System pressure to less than 1000 psig within-the fol-lowing 6 hours. 3) The volume weighted average boron concentration of the tbme - limiting accumulators 1500 ppm or less, return the volume weighted average boron concentration of the three limiting acc oulators to greater than 1500 ppm and enter ACTION c.2 within 1 hour of the low boron determination or be in HOT STANDBY within the next 6 hours and reduce Reactor Coolant System pressure to less than 1000 psig within the following l 6 hours. SURVEILLANCE REQUIREMENTS 4.5.1 Each cold leg injection accumulator shall be demonstrated OPERABLE: [ a. At least once per 12 hours by: 1) Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and 2) Verifying that each cold leg injection accumulator isolation valve is open. b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 75 gallons by verifying the boron concentration of the accumulator solution; At least once per 31 days when the Reactor Coolant System pressure c. is above 2000 psig by verifying that power is removed from the isolation valve operators on Valves NI54A, NI658, NI76A~,~and NI888 and that the respective circuit breakers are padlocked; and d. At least once per 18 months by verifying that each cold leg injection accumulator isolation valve opens automatically under each_of the following conditions:** 1) When an actual or a simulated Reactor Coolant System pressure signal exceeds the P-11 (Pressurizer Pressure Block of Safety-Injection) Setpoint, and 2) Upon receipt of a Safety Injection test signal.

- This surveillance need not be performed until prior to entering HOT STANDBY l

following the Unit 1 refueling. l CATAWBA - UNITS 1 & 2 3/4 5-2 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

EMERGENCv CORE COOLING SYSTEMS SURVEILLA,N,CE REQUIREMENTS (Continued) 4.5.1 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 a. OPERATIONAL TEST, and b. At least once per 18 months by the performance of a CHANNEL.._ CALIBRATION. l l l-9 CATAWBA - UNITS 1 & 2 3/4 5-3 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

Page 3/4 5-4 is intentionally blank, l i CATAWBA - UNITS 1 & 2 3/4 5-4 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2) - --- - - ~ - - - - - ~ - - - -

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b) With a simulated or actual Reactor Coolant System pressure signal less than or equal to 660 psig the intetlocks will cause the valves to automatically close. 2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.'" At least once per 18 months, during shutdown, by:** e. 1) Verifying,that each automatic valve in the flow path actuates to its correct position on Safety Injection and Containment Sump Recirculation test signals, and 2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection test signal: a) Centrifugal charging pump, b) Safety Injection pump, and c) Residual heat removal pump. f. By verifying that each of the following pumps develops the indicated differential pressure when tested pursuant to Specification 4.0.5: 1) Centrifugal charging pump 1 2380 psid, 2) Safety Injection pump 1 1430 psid, and 3) Residual heat removal pump 1 165 psid. g. By verifying the correct position of each electrical and/or mechanical stop for the following ECCS throttle valves: 1) Within 4 hours following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and 2) At least once per 18 months. Centrifugal Charging Pump Injection Throttle Safety Injection Throttle Valve Number Valve Number NI-14 NI-164 NI-16 NI-166 HI-18 NI-168 NI-20 NI-170 "" This surveillance need not be performed until prior to entering HOT SHUTDOWN following the Unit One first refueling. CATAWBA - UNITS 1 AND 2 3/4 5-7

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) h, By performing a flow balance test, during shutdown, following com-pletion of modifications to the ECCS subsystems that alter the subsystem flow characteristics and verifying that: 1) For centrifugal charging pump lines, with a single pump running: a) The sum of the injection line flow rates, excluding,,.the highest flow rate, is greater than or equal to 345 gpm, l and b) The total pump flow rate is less than or equal to 565 gpm. 2) For Safety Injection pump lines, with a single pump running: a) The sum of the injection line flow rates, excluding the highest flow rate, is greater than or equal to 450 gpm, l and b) The total pump flow rate is less than or equai to 660 gpm. 3) For residual heat removal pump lines, with a single pump running, the sum of the injection line flow rates is greater than or equal to 3648 gpm. l 4 CATAWBA - UNITS 1 & 2 3/4 5-8 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

3/4.2 POWER DISTRIBUTION LIMITS (Unit 1) l BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) maintaining the calculated DNBR in the core greater than or equal to design limit DNBR during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. In addition,.)(miting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA anal, es are met and the ECCS acceptance criteria are not exceeded. I The definitions of certain hot channel and peaking factors as used in these specifications are as follows: F (X,Y,Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat q flux on the surface of a fuel rod at core elevation Z divided by the l average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; F H(X,Y) Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of l the integral of linear power along the rod with the highest integrated power to the average rod power. K(z) is defined as the normalized F (X,Y,Z) limit for a given core height. 9 l 3/4.2.1 AXIAL FLUX OIFFERENCE-Unit 1 The limits on AXIAL FLUX DIFFERENCE (AFD) specified in the CORE OPERATING LIMITS REPORT (COLR) ensure that the F (X,Y,Z) and the F3g(X,Y) limits are q not exceeded during either normal operation or in the event of xenon redistrib-ution following power changes. The AFD envelope specified in the COLR has been adjusted for measurement uncertainty. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, AND NUCLEAR ENTHALPY RISE l HOT CHANNEL FACTOR (Unit 1) l The limits on heat flux hot channel factor, 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 ECCS ac-ceptance criteria are not exceeded. The peaking limits are specified in the CORE OPERATING LIMITS REPORT (COLR) per Specification 6.9.1.9. The heat flux hot channel factor and nuclear enthalpy rise hot channel factor are each 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: 1 I 4 CATAWBA - UNITS 1 & 2 8 3/4 2-1 Amendment No. 86 (Unit 1) l Amendment No. 80 (Unit 2) i

POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR, AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Unit 1) (Continued) 1 Control rods in a single group move together with no individual rod a. insertion differing by more than 2 12 steps, indicated, from the group demand position; b. Contro! rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; The control rod insertion limits of Specifications 3.1.3.5 and c. 3.1.3.6 are maintained; and The axial power distribution, expressed in terms of AXIAL FLUX d. DIFFERENCE, is maintained within the limits. FAH(X,Y) will be maintained within its limits provided Conditions a. above are maintained. l through d. The limits on the nuclear enthalpy rise hot channel factor, FAH(X,Y), are specified in the COLR as Maximum Allowable Radial Peaking limits, obtained by dividing the Maximum Allowable Total Peaking (MAP) limit by the axial peak [ AXIAL (X,Y)] for location (X,Y). By definition, the Maximum Allowable Radial Peaking limits will, for Mark-BW fuel, result in a ONBR for the limiting tran-sient that is equivalent to the DNBR calculated with a design FAH(X,Y), value of 1.55 and a limiting reference axial power shape. The Mark-BW MAP limits may be applied to 0FA fuel, provided an appropriate adjustment factor is applied to provide equivalence to a 1.49 design FAH(X,Y), for the OFA. This is reflected in the, MAP limits specified in the COLR. The relaxation of FAH(X,Y), as a func-tion of THERMAL POWER allows changes in the radial power for all permissible control bank insertion limits. This relaxation is implemented by the application of the following factors: k = [1 + (1/RRH) (1 - P)] where k = power factor multiplier applied to the MAP limits P = THERMAL POWER / RATED THERMAL POWER RRH is given in the COLR FQ"(X,Y,2) and FAHR (X,Y) are measured periodically, and comparisons to N the allowable limit are made to provide reasonable assurance that the limiting criteria will not be exceeded for operation within the Technical Specification limits of Sections 2.2 (Limiting Safety Systems Settings), 3.1.3 (Movable Control Assemblies), 3.2.1 (Axial Flux Difference), and 3.2.4 (Quadrant Power Tilt Ratio). A peaking margin calculation is performed to provide a basis for decreasing the width of the AFD and f(AI) limits and for reducing THERMAL POWER. 8 CATAWBA - UNITS 1 & 2 B 3/4 2-2 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2) l

POWER DISTRIBUTION LIMITS BASES h' EAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FA (Unit 1) (Continued) When an FQ"(X,Y,Z) measurement is obtained in accordance with the surveil-lance reauirements of Specification 4.2.2, no uncertainties are applied to the measured peak; the required uncertainties are included in the peaking limit. When FQ"(X,Y,Z) is measured for reasons other than meeting the requiremerits of Specification 4.2.2, the measured peak is increased by the radial-local peaking factor to convert it to a local peak. Allowances of 5% for measurement uncer-tainty and 3% for manufacturing tolerances are then applied to the measured peak. When an FAHR (X,Y) measurement is obtained, regardless of the reason, no uncertainties ate applied to the measured peak; the required uncertainties are included in the peaking limit. 3/4.2.4 QUADRANT POWER TILT RATIO (Unit 1) The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation, The limit of 1.02, at which corrective action is required, provides ONB and linear heat generation rate protection with x y plane power tilts. A peaking increase that reflects a QUADRANT POWER TILT RATIO of 1.02 is included in the generation of the AFD limits. 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 correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on F (X,Y,Z) is reinstated by I 9 reducing the maximum allowed power by 3% for each percent of tilt in excess of 2L I For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. 3/4.2.5 ONB PARAMETERS-(UNIT 1) The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a CATAWBA - UNITS 1 & 2 8 3/4 2-3 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

POWER DISTRIBUTION LIMITS BASES-3/4.2.5 DNB PARAMETERS-(UNIT 1) (Continued)- design limit DNBR throughout each analyzed transient. As noted on Figure 3.2-1, Reactor Coolant System flow rate and THERMAL POWER may be " traded off" against one another (i.e., a low measured Reactor Coolant System flow rate is acceptable if the THERMAL POWER is also low) to ensure that the calculated DNBR will not be below the design DNBR value. The relationship defined on Figure 3.2-2 remains valid as long as the limits placed on the nuclear enthalpy rise hot cha"nnel N factor, F aH, in Specification 3.2.3 are maintained. The indicated T,yg value and the indicated pressurizer pressure value correspond to analytical limits of 59a.8*F and 2205.3 psig respectively, with allowance for measurement uncer-tainty. When Reactor Coolant System flow rate is measured, no additional allowances are necessary prior to comparison with the limits of Figure 3.2-1 since a measurement error of 2.1% for Reactor Coolant System total flow rate has been allowed for in' determination of the design DNBR value. The measurement error for Reactor Coolant System total flow rate is based upon performing a precision heat balance and using the result to calibrate the Reactor Coolant System 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 nonconservative manner. 'Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is included in Figure 3.2-1. Any fouling which might bias the Reactor Coolant System 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 Reactor Coolant System flow rate measurement or-the venturi shall be cleaned to eliminate the fouling, The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. Indica-tion instrumentation measurement uncertainties are accounted for in the limits provided in Table 3.2-1. CATAWBA - UNITS 1 & 2 B 3/4 2-4 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2) ~

3/4.2 POWER DISTRIBUTION LIMITS (Unit 2) l BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (1) maintaining the calculated DNBR in the-core greater than or equal to design limit DNBR during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria. Inaddition,lAntting 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 limit of 2200 F is not exceeded. The definitions of certain hot channel and peaking factors as used in these specifications are as follows: F (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat 9 flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; Fh Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. 3/4.2.1 AXIAL FLUX DIFFERENCE (Unit 2) l The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (2) upper 9 bound envelope of the F limit specified in the CORE-0PERATING LIMITS REPORT (COLR) times the normalized axial 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 reds may be positioned within the core 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 target flux' difference obtained_under these conditions divided by the fraction of RATED THERMAL POWER is-the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences-for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux = difference value is necessary to reflect core burnup considerations. 4 -CATAWBA - UNITS 1 & 2 B 3/4 2-5 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

. = _ POWER DISTRIBUTION LIMITS BASES ND At power levels below APL , the limits on AFD are defined in the COLR, i.e., that defined by the RAOC operating procedure and limits. These limits were calculated in a manner such that expected operational transients, e.g., load follow operations, would not result in the AFD deviating outside of those limits. However, in the event such a deviation occurs, the short period of time allowed-outside of the limits at reduced power levels will not result in signi-ficant xenon redistribution such that the envelope of peaking factors wou,ld change sufficiently to prevent operation in the vicinity of the APLND level. pgwer ND At power levels greater than APL , two modes of operation are permis-sible; 1) RAOC, the AFD limits of which are defined in the COLR, and 2) Base load operation, which is defined as the maintenance of the AFD within a COLR specified band about a target value. The RA00 operating procedure above ND APL is the same as that defined for operation below APL"O However, it is possible when following extended load following maneuvers that the AFD limits may result in restrictions in the maximum allowed power or AFD in order to guarantee operation with F (z) less than its limiting value. To allow operation 9 at the maximum permissible value, the Base Load operating procedure restricts l l 1 i CATAWBA - UNITS 1 & 2 B 3/4 2-6 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

POWER DISTRIBUTION LIMITS BASES the indicated AFD to relatively small target band and power swings (AFD target band as specified in the COLR, APL"O L $ power 5 APL or 100% Rated Thermal Power, whichever is lower). For Base Load operation, it is expected that the Units will operate within the target band. Operation outside of the target band for the short time period allowed will not result in significant xenon redistribution such that the envelope of peaking factors would change sufficiently to.pr.ohibit continued operation in the power region defined above. To assure there is no residual xenon redistribution impact f om past operation on the Base Load operation, a 24 hour waiting period at a power level above APLND and allowed by RA0C is necessary. During this time period load changes and rod motion are restricted to that allowed by the Base load procedure. After the waiting period extended Base Load operation is permissible. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are:

1) outside the allowed AI power operating space (for RA0C operation), or 2) outside the allowed AI target band (for Base Load operation).

These alarms are active when power is greater than:

1) 50% of RATED THERMAL POWER (for RA0C operation), or

2) APLND (for Base Load operation).

Penalty deviation minutes for Base Load operation are not accumulated based on the short period of time during which operation outside of the target band is allowed. 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Unit 2) l The limits on heat flux hot channel factor, coolant 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 acceptance criteria limit. These limits are specified in the CORE OPERATING LIMITS REPORT per-Specification 6.9.1.9. 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: Control rods in a single group move together with no individual rod a. insertion differing by more than i 12 steps, indicated, from the group demand position; b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; s CATAWBA - UNITS 1 & 2 B 3/4 2-7 Amendment No.86 Unit 1) l Amendment No.80 ((Unit 2)

POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLE ENTHALPY RISE HOT CHANNEL iACTOR (Unit 2) (Continued) l The control rod insertion limits of Specifications 3.1.3.5 and c. 3.1.3.6 are maintained; and d. The axial power distribution, expressed in terms of AXIAL FLUX, DIFFERENCE, is maintcined within the limits. F will be maintained within its limits provided Conditions a, through d. g above are maintained. As noted on the figure specified in the CORE OPERATING LIMITS REPORT (COLR), Reactor Coolant System flow rate and F may be " traded g of f" against one another (i.e., a low measured Reactor Coolant System flow rate N is acceptable if the measured F is also low) to ensure that the calculated H DNBR will not be below the design DNBR value. The relaxation of F as a N 6H 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 the figure specified in the COLR, accounts for F less than or equal to the F limit specified H in the COLR. This value is used in the various accident analyses where F influences parameters other than DNBR, e.g., peak clad temperature, and thus g is the maximum "as measured" value allowed. The rod bow penalty as a function of burnup applied for F is calculated with the methods described in WCAP-8691, q Revision 1, " Fuel Rod Bow Evaluation," July 1979, and the maximum rod bow penalty is 2.7% DNBR. Since the safety analysis is performed with plant-specific safety DNBR limits compared to the design DNBR limits, there is sufficient thermal margin available to offset the rod bow penalty of 2.7% DNBR. The hot channel factor F (z) is measured periodically and increased by a cycle and height dependent power factor appropriate to either RA0C or Base Load operation, W(z) or W(z)BL, to provide assurance that the limit on the hot channel factor, F (z), is met. W(z) accounts for the effects of normal oper-9 ation transients and was determined from expected power control maneuvers over l the full ~ range of burnup conditions in the core. W(z)BL accounts for the more restrictive operating limits allowed by Base Load operation which result in less severe transient values. The W(z) function for normal operation and the W(Z)BL function for Base Load Operation are specified in the CORE OPERATING LIMITS REPORT per Specification 6.9.1.9. 1 i CATAWBA - UNITS 1 & 2 B 3/4 2-8 AmendmentNo.g(Unit 1) Amendment No. (Unit 2) l

POWER DISTRIBUTION LIMITS i BASES HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR D THALPY RISE HOT CHANNEL FACTOR (Unit 2) (Continued) When Reactor Coolant System flow rate and FNg are measured, no additional allowances cre necessary prior to comparison with the limits of the figure specified in the COLR. Measurement errors of 2.1% for Reactor Coolant System total flow rate and 4% for F have been allowed for in determination of,4he H design DNBR value. The measurement error for Reactor Coolant System total flow rate is based upon performing a precision heat balance and using the result to calibrate the Reactor Coolant System 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 nonconservative manner. Therefore, a penalty of 0.1% for undetected-fouling of the feedwater venturi is included in the figure specified in the COLR. Any fouling which might bias the Reactor Coolant System flow rate measurement greater than 0.L% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before per-forming subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be _ quantified and compensated for in the Reactor Coolant System flow rate measurement or the venturi shall be cleaned to eliminate the fouling. The 12-hour periodic surveillance of indicated Reactor Coolant System flow is sufficient to detect only flow degradation which could lead to opera-tion outside the acceptable region of operation specified on the figure spec-ified in the COLR. 3/4.2.4 QUADRANT POWER TILT RATIO (Unit 2) l The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion _ satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation. The_ limit of 1.02, at which corrective action is required, provides DN8 and' linear heat generation rate protection with x y plane power tilts. A limit of 1.02 was selected to provide an allowance for the uncertainty associated

the indicated power-tilt.

?-hour time allowance for operation with-a tilt condition greater but less than 1.09-is provided to allow identification and correction md or misaligned control rod. In the event such action does not che tilt, the margin for uncertainty on F is reinstated by reducing u g the maximum allowed power by 3% for aach percent of_ tilt in excess of 1. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore y detector is inoperable, the movable incure detectors are used to confirm that L - the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore i CATAWBA - UNITS 1 & 2 8 3/4 2-9 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

POWER DISTRIBUTION LIMITS BASES QUADRANT POWER TILT RATIO (Unit 2) (Continued) flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. The normal locations are C-8, E-5, E-11 H-3, H-13, L-5, L-11, N-8. Alternate locations are available if any of the normal locations are unavailable. 3/4.2.5 DNB PARAMETERS (Unit 2) The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a design limit DNBR throughc.at each analyzed transient. The indicated T,yg value and the indicated pressurizer pressure value correspond to analytical limits of 594.8 F and 2205.3 psig respectively, with allowance for measurement uncertainty. The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. Indica-tion instrumentation measurement uncertainties are accounted for in the limits provided in Table 3.2-1. CATAWBA - UNITS 1 & 2 8 3/4 2-10 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

.~ - -. - -. - - - - ADMINISTRATIVE CONTROLS -SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT (Continued) The Radioactive Effluent Release Reports shall include a list and descrip-tion of unplanne _ releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period. The Radioactive Effisnt Release Reports shall include any changes made during the reporting oeriod t? the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALC'IL% TIM MANUAL (00CM), as well as a listing of new locations for dose calc 91ap ons a Q /or environmental monitoring identified by the land use census purwin', to Spacification 3.12.2. MONTHLY OPERATING REPOR13

6. 9.1. 8 Routine reports of operating statistics and shutdown experience, in-cluding documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the NRC in accordance with 10 CFR 50.4, no later than the 15th of each month following the calendar month covered by the report.

CORE OPERATING LIMITS REPORT

6. 9.1. 9 Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

1. Moderator Temperature Coefficient BOL and E0L limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, 2. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, 3. Control Bank Insertion Limits for Specification 3/4.1.3.6, NP 4. Axial Flux Difference Limits, target band *, and APL for l Specification 3/4.2.1, RTP 5. Heat Flux Hot Channel Factor, F K(Z), W(Z)**, APLND" and W(Z)BL forSpecification3/4.2.3,a,nd 6. Nuclear Enthalpy Rise Hot Channel-Factor, FAHRL*** or, FRTP AH l and Power Factor Multiplier,-MFAH , limits for Specifica-tion 3/4.2.3. The' analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in: 1. WCAP-9272-P-A, " WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY," July 1985 (W Proprietary). (Methodology for Specifications 3.1.1.3 - Moderator Temperature Coefficient, 3.1.3.5 - Shutdown Bank Insertion Limit, 3.1.3.6 - Control-Bank Insertion Limits, 3.2.1 - Axial Flux ND

Reference 5-is not applicable to target band and APL. ND
- References 5 and 6 are not applicable to W(Z), and APL

,andW(p.

- - Reference 1 is not applicable to FAHR'.

RTP

- - - Reference 5 is not applicable to F and MFg.

CATAWBA - UNITS 1 & 2 6-19 Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2)

ADMINISTRATIVE CONTRO o CORE OPERATING LIMITS REPORT (Continued) Difference, 3.2.2 - Heet Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Factor.) 2. WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFSET CONTROL FQ SURVEILLANCE TECHNICAL SPECIFICATION," June 1983 (W Proprietary). (Methodology for Specifications 3.2.1 - Axial Flux Difference (Relaxed Axial Offset-Control) and 3.2.2 - Heat Flux Hot Channel Factor (W(Z) surveillance requirements for F Methodology.) q 3. WCAP-10266-P-A Rev. 2, "THE 1981 VERSION OF WESTINGHOUSE CVALUATION MODEL USING BASH CODE," March 1987, (W Proprietary). (Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.) 4. BAW-10152-A, " NOODLE - A Multi-Dimensional Two-Group Reactor Simulator," June 1985. (Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.) 5. BAW-10163P-A, " Core Operating Limit Methodology for Westinghouse-Designed PWR's," June 1989. ('M'ethodology for Specifications 3.1 3.5 - Shutdown Rod Insertion Limits, 3.1.3.6 - Control Bank Iasertion Limits, 3.2.1 - Axial Flux Difference, 3.2.2 - Heat Flux Hot Channel Factor, and 3.2.3 - Nuclear Enthalpy Rise Hot Channel Fattor.) 6. BAW-10168P, Rev.1, "B&W Loss of-C dar.t Accident Evaluation Model for Recirculating Steam Generator Plant.,,'_' September,1989. (Methodology for Specification 3.2.2 - Heat Flux Hot. Channel Factor.) The core operating limits shall be determineo-so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic-limits,-ECCS limits,_ nuclear limits such as shutdown margin, and transient and_ accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT, including any mid-cycle _ revisions or supplements thereto, shall be pro 91ded upon issuance, for each reload cycle,. to the NRC in accordance with 10.CFR 50.4. u i: \\ l CATAWBA - UNITS 1 & 2 6-19a Amendment No. 86 (Unit 1) Amendment No. 80 (Unit 2) . - _ - -.}}

ML20073R450

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